REVIEW OF THE - Welcome to NewMexico.gov Isolation Pilot...evaluations on the waste hoist system...
Transcript of REVIEW OF THE - Welcome to NewMexico.gov Isolation Pilot...evaluations on the waste hoist system...
REVIEW OF THE FINAL SAFETY ANALYSIS REPORT (DRAFT), W E WASTE ISOLATION PILOT PLANT, DECEMBER 1988
Environmental Evaluation Group 7007 Wyoming Boulevard NE, Suite F-2
Albuquerque, New Mexico 87109
May 1989
CONTENTS
Title Page
FOREWORD.......................................... i
STAFF............................................. ii
SUMMARY........................................... 1
Comments on CHAPTER I............................. 8
Comments on CHAPTER I A . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
Comments on CHAPTER 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
Comments on CHAPTER 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
Comments on CHAPTER 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
Comments on CHAPTER 5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
Comments on CHAPTER 6 (including APPENDIX 6A) ..... 4 6
Comments on CHAPTER 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 0
Comments on CHAPTER 8 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 60
Comments on CHAPTER 9 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 6
Comments on CHAPTER lo............................ 69
Comments on CHAPTER ll............................ 7 3
Comments on CHAPTER 12............................ 7 4
Appendix: October 1988 Comments.................. 7 6
The purpose of the Environmental Evaluation Group (EEG) is to
conduct an independent technical evaluation of the Waste
Isolation Pilot Plant (WIPP) Project to ensure protection of the
public health and safety and the environment. The WIPP Project,
located in southeastern New Mexico, is being constructed as a
respoitory for permanent disposal of transuranic (TRU)
radioactive wastes generated by the national defense programs.
The EEG was established in 1978 with funds provided by the U. S.
Department of Energy (WE) to the State of New Mexico. Public
Law 100-456, the National Defense Authorization Act, Fiscal Year
1989, Section 1433, assigned EEG to the New Mexico Institute of
Mining and Technology and provided for continued funding from
W E through Contract DE-AC04-79AL10752.
EEG performs independent technical analyses of the suitability of
the proposed site; the design of the repository, its planned
operation, and its long-term integrity; suitability and safety of
the transportation systems; suitability of the Waste Acceptance
Criteria and the generator sites' compliance with them; and
related subjects. These analyses include assessments of reports
issued by the DOE and its contractors, other federal agencies and
organizations, as they relate to the potential health, safety and
environmental impacts from WIPP. EEG also performs environmental
monitoring for background radioactivity in air, water, and soil,
both on-site and in surrounding communities.
" Robert H. Neil1
Director
STAFF
Sally C. Ballard, Environmental Technician
William T. Bartlett, Ph.D., Health Physicist
James K. Channell, Ph.D., P.E., CHP, Sr. Environmental Engineer
Lokesh Chaturvedi, Ph.D., Deputy Director h Engineering Geologist
Laura H. Connolly, M.A., Librarian/Editor
Anthony F. Gallegos, Ph.D., Sr. Performance Assessment Specialist
Jim W. Kenney, M.S., Environmental Scientist
C. Robert McFarland, B.S., Sr. Quality Assurance Engineer
Robert H. Neill, M.S., Director '
Kevin J. Shenk, M.S., Health Physicist
Mary Stetton, B.A., Administrative Secretary
Susan C. Stoltum, Administrative Secretary
Georgina M. -Temple, Secretary
Brenda J. West, B.A., Administrative Officer
These are the Environmental Evaluation Group's comments
on the December 1988 Draft of the W E Final Safety Analysis
Report (FSAR) on WIPP which was received by EEG on February
13, 1989. Previously we commented (October 14, 1988 letter
from the Director of EEG to the DOE WIPP Project Manager) on
the earlier Draft FSAR, the initial 1980 Safety Analysis
Report (SAR), and all nine subsequent amendments to the SAR.
... EEG1s comments on the SAR and its amendments are summarized : t ...,
r' ,, \, i n EEG-29, Marshall S. Little, nEvaluation of the Safety
/ 4." : , ' , ji Ti; ": ' :Analysis Report for the Waste Isolation Pilot Plant Pr~ject,~' ,, &? 2 i .:: ..:, ? \.;,: , ; , I " , g . M & h a y 1985.
ii% -,*1.- Y
EEG8s October 1988 comments were extensive and indicated
there was much we believed needed to be corrected, amplified,
and included in the FSAR. The current comments are equally - extensive and we believe a significant amount of work still
needs to be done by DOE to produce an acceptable FSAR. Many
comments are repeated from our October 1988 review because
the December 1988 Draft either failed to respond to all of
our comments, or responded inadequately. Additional comments
on the current draft involve the hazardous wastes analyses in
Chapter 6 and 7 which are in the Draft FSAR for the first
time. A few of the comments address issues not covered in
our October 1988 review but which may have been raised
previously in EEG's comments on the various SAR amendments or
in other WIPP Project Office ( W W ) Reports.
Another point needs to be made. The FSAR does not
contain all of the detailed information and procedures
necessary to determine operational readiness of the WIPP
facility. In fact, it is apparent from the detailed description in the Working Agreement to the July 1, 1989,
Consultation and Cooperation Agreement (see Appendix B,
1
Working Agreement, Revision 1, March 23, 1983, Article IV, K.
Operations) that the review of the FSAR is intended to be
only one of the milestones under the Operations Key Event and
not the sole criteria for determining operational readiness.
EEG has concluded, from participation in the DOE Preopera-
tional Readiness Appraisal in March 1989, that there are a
significant number of outstanding items. We have made no
attempt to address all of these concerns in our FSAR
comments because they don't appear to be pertinent to the
document and because we intend to make our operational
readiness decision separate from the FSAR.
Our more important comments are summarized below in the
, . -- order the items are addressed in the Draft FSAR. The
ignificance-we attach to each should be apparent from our
iscussion. More detail is provided on these issues under
ndividual chapter comments.
. . . '. -.
1. Since the FSAR does not include the long-term risk
assessment required by EPA in their disposal standards for
TRU waste (40 CFR 191, Part B Performance Assessment), the Safety Analysis Report only applies to the five-year
demonstration phase of the project. This should be clearly
stated in Chapter 1 of the FSAR. There must then be a
supplement to the FSAR prior to a disposal phase of the WIPP Project. Also, the supplement would need to contain the
operational safety requirements for handling RH-TRU waste.
2. The Second Modification to the July 1, 1981,
"Agreement for Consultation and Cooperationn on WIPP by the
State of New Mexico and the U.S. Department of Energy
requires the FSAR to "document DOE'S ability to comply with
the provisions of Subpart A of...(40 CFR 191)." This Draft
FSAR does not explicitly address Subpart A compliance. Explicit documentation must be provided.
3. EEG strongly objects to the exclusion of calcula-
tions for the hoist drop scenario (C8) on the grounds that it
is not a credible event. We have never believed that
failure mode analysis can be relied on to prove that low
probability accidents will not occur. The history of
evaluations on the waste hoist system support this skep-
ticism. The DOE WIPP Project Office published a report in
1985 showing the probability of a catastrophic accident was
less than 1E-07 per year. Then in 1987, as part of the
Operational Readiness Review, a W E study indicated the
probability to be 1E-03 per year. But we are assured (on
page 1A.4-5) that this will be corrected and the event will
still be incredible. Implicit in this assurance is the
belief that every possible failure sequence has now been
recognized and correctly evaluated. EEG does not share this
conclusion and believes calculations for this scenario should
be included in the FSAR.
4. We consider the assumed low failure rate (1E-04) of
the exhaust filtration system claimed on page 1A.4-6 to be -
"" unproven. For one thing, the system did not work properly / "
k * during a scheduled drill during the March 1989 Preoperational ' % * * , ) I % . ' ~ +
, Appraisal Audit. Secondly, even if this portion of the I- system performed perfectly, the Continuous Air Monitors
-, (CAMS) would also have to perform adequately with the
required sensitivity in order to signal the switch to the
filtration mode. As mentioned below, we also consider the
CAM system to be unproven. We believe a great deal of effort
needs to go into proving the capability and insuring the
reliability of both systems.
5 . The FSAR should specify in as much detail as possible the volumes, curies, and distribution within both
CH-TRU and RH-TRU containers and the totals. Also, possible
ranges and uncertainties in these estimates should be - discussed directly. All of these parameters have a bearing
on evaluating different aspects of safety at WIPP and are
required (by Article VI of the Consultation and Cooperation
Agreement) to be included. (See comment in Chapter 3.)
6. Appendix 6A, which contains the methodology
necessary for evaluating airborne radionuclide concentrations
from routine operations, contains serious flaws. The
corresponding radionuclide concentrations and doses in
Chapter 6 are incorrect. Appendix 6A and related portions of
2hapter 6 need to be completely redone.
7. The CAMs located in the Waste Handling Building,
underground,-and in the Exhaust Filtration Building are vital
to the protection of workers and to the warning of environ-
-. _,.- mental releases. The ability of these instruments to detect
airborne radionuclide concentrations with the required
degree of sensitivity has not yet been proven. The adequacy
of the CAMs must be established by the WW and verified by an
outside peer review group, including EEG, before wastes can
be brought to WIPP. (See Chapter 6 comments.)
8. Several of our comments on Chapter 6 refer to
concerns about the WIPP Operational Health Physics Program.
Since the FSAR does not address this Program in a comprehen-
sive manner, these comments will not respond in detail.
However, EEG's serious reservations about the present status
of this Program were provided to DOE in our April 7, 1989,
comments on the WIPP Phase I1 Preoperational Appraisal.
9. Analytical samples for both high- and low-level
counting rooms should not be prepared in the same preparation
room. Also, routine (as well as incident) bioassays must be
carried out on radiation workers. (See Chapter 6 comments.) A
10. The use of a 1000 PE-Ci upper limit for individual
waste containers at WIPP is unacceptable to EEG. Even with a
somewhat lower limit it may still be necessary to impose
operational restrictions on high-curie drums. (See our
comments under Chapter 7.)
11. The potential doses calculated in Chapter 7 to
radiation workers from accidents involving CH-TRU waste
handling are unreasonably low because the assumptions include
., only an average PE-Ci quantity in a drum and because (in the , , ' ' . C2 and C3 scenarios) the forklift operator is not considered
to be exposed. It should be recognized that very high
occupational doses are possible and that operational
restrictions need to be employed to minimize them.
12. It is appropriate to include the safety aspects of
the non-radiological hazardous waste component coming to
WIPP. There are some numerical inconsistencies or am-
biguities in this draft (see our comments on Chapters 6 and
7) which should be corrected in the next draft. Our
principal observations on the methodology and assumptions
are:
(a) the exclusion of all chemicals that represent less than 1% of the hazardous waste constituents (by weight) may
not be conservative, because toxicities can vary over several
orders of magnitude, and should be reconsidered;
(b) the assumption that average concentrations of Rocky
Flats Plant waste are conservative averages for the entire
system has not been adequately explained:
(c) the atmospheric dispersion models used give drasti-
cally different results than the ones used for radionuclides;
and
(d) assumptions about zero mobility of lead and any
other hazardous chemical (except VOC's in head space gas)
5
following accidents is non-conservative and inconsistent with -. assumptions and observations about loss of transuranics
(which should be as immobile as virtually all chemicals) from
damaged containers.
Because of the extensive re-writing needed on the
hazardous waste sections in Chapters 6 and 7 we will not try
to reach a conclusion at this time about the potential
hazard.
13. The FSAR takes credit in Chapter 8 for a Peer
Review Panel providing assurance on suitability of WIPP as a
repsitory. Since the Department has never involved EEG with
any of the Peer Review Panels, nor provided us with the
agenda, minutes, or recommendations, we believe that the
mmittees do not provide credibility as stated in the FSAR,
t in actuality detract from it. In order to take credit,
G must be involved.
14. It is noted that all references to backfilling the
waste storage rooms have been deleted from the FSAR. The
FSAR should clearly state whether backfilling will be done
during the experimental phase, when the decision on backfill
will be made, and the probable final backfill design during
operation. (See Chapter 1 comments.)
15. The Operational Safety Review (Chapter 10) lacks
sufficient detail to permit us to evaluate the operational
safety of WIPP. EEGvs specific comments describe the areas
in which extensive expansion and revision are needed.
16. A Design Basis Accident (DBA) assessment addressing the requirements of W E Order 6430.1 and the guidelines to "A
Guide to Radiological Accident Consideration for Siting and
Design of W E Non-Reactor Nuclear Facilities," LA-10294-AC, -
should be performed and summarized in Chapter 7.
17. This draft of the FSAR evaluates only the proposed
disposal mode waste emplacement procedure. Yet, the FSAR is
fully applicable to only the (approximately) Five-Year
Performance Assessment and Operational Demonstration Test
Phase which is not evaluated in any manner. The FSAR should
contain analyses of the operations required during the Test
Phase. The analysis should include the period at the end of
the Test Phase when the wastes at WIPP must be either
permanently emplaced or retrieved and shipped elsewhere.
(See our comments in Chapters 6 and 12.)
CBAPmeR1 Introduction and General Description
A. General Comments
1. This chapter, including Appendix lA, has several
significant improvements, including a more comprehensive list
of pertinent references and an updating of general scien-
tific, technical and physical descriptions. The WPO has
made significant responses to EEG1s previous comments and
recommendations. The chapter provides a good introduction to
the FSAR, and there are only a few areas where further
changes arexecommended. These are discussed in the detailed
comments below.
2. After reviewing the entire PSAR, it was noted that
it does not contain the performance assessments required by
40 CFR 191 Part B. Therefore, the safety analyses appear to
be limited to the five-year pilot plant phase of the WIPP
operations. This fact should be clearly stated in Chapter 1.
3 . Chapter 1 should include a brief discussion of the
status of the potash leases which cover part of the WIPP
site, and indicate how and when W E plans to address this
problem.
4 . The decision to dispose of the wastes at WIPP or
retrieve them will not be based on the results of the Five-
Year Pilot Phase as described in the text, but should be
based on the ability to meet the EPA Standards for disposal.
B. Detailed Comments
1. Section 1.1, Introduction, Page 1.1-2. The second
paragraph on this page refers to the WIPP facility as a islow
hazard facility," namely, one *@which presents minor onsite
and negligible offsite impacts to people of the environment."
This conclusion is contrary to WE/AL Order 5481.1B.
According to this Order, a hazard facility involves only
"hazards of a type and magnitude routinely encountered and
accepted by the public." As pointed out in the FSAR, the
WIPP is a first-of-a-kind facility and, pursuant to W E Order
5480.5, it is a nuclear facility. Therefore, the hazards of
a nuclear facility should not be considered as a type
"routinely encountered and accepted by the p u b l i ~ . ~ DOE/AL Order 5481.lB also States that all nuclear facilities shall
be supported by a SAR. The second paragraph on page 1.1-2
also states that this classification of WIPP as a low hazard
facility is in'accord with Chapter I1 of WE/AL Order
5481.1B. Chapter I1 discusses "Operational Safety
Requirementsw and provides no guidance whatsoever for
classification of nuclear facilities.
On page 1.1-4, the third paragraph discusses experiments
and operational demonstrations needed to reach a decision
, regarding the permanent isolation of wastes at WIPP. Such
experiments and demonstrations are not described in the FSAR,
therefore it does not seem appropriate to refer to them in
Chapter 1 until such time as they become a part of the FSAR.
Also, the latter part of this paragraph discusses the EPA
Standard, 40 CFR 191, and leaves the impression that WIPP
compliance with the standard is delayed because the standards
were vacated and remanded by the courts back to EPA for
reconsideration. It should be added that W E and the State
of New Mexico have formally agreed that DOE will proceed to
demonstrate compliance with the vacated standard.
- On page 1.1-7, there is a statement that technical data,
unless it directly supports the Safety Analysis of recent
facility modifications, is current through December 1986.
Please clarify why technical data on site characterization
collected in 1987 and 1988 is not included.
At the bottom of page 1.1-7 and top of page 1.1-8, there
is a discussion of the plans to prepare a Supplemental
Environmental Impact Statement. Among the reasons which
should be added for the need for the SEIS is to summarize the
considerable amount of information acquired between 1980 and
1989. This information may change the assessment of the site
for the mission of WIPP. For example, concerning the Castile
brine reservoirs, FEIS (p. 9-134) stated, "...brine pockets
of the size assumed in this example are extremely unlikely
/" a near the repository ..." In 1981, WIPP-12 borehole, located at the edge of the repository, was deepened and a brine -
," reservoir was encountered which was estimated to contain 17
million barrels of brine, 2 million barrels more than the
amount assumed in the FEIS example. After the location of
the repository was moved 1.25 miles to the south geophysical
surveys performed over the new location showed, "...brine
appears to be present 250 meters below portions of the waste
panel horizon..." (DOE letter to EEG, 12/29/87). Many such
examples of new facts, revised concepts and updated data will
need to be addressed in the SEIS.
2. Section 1.1.2, Mission, Page 1.1-11. The Mission
Statement emphasizes the research and development aspect of
WIPP and mentions the possibility that wastes will not be
permanently disposed of at the site. Because of this pos-
sibility, the FSAR should indicate where retrieved Waste
would be sent for storage or disposal.
3. Section 1.1.2, Mission, Page 1.1-11. The Mission
of WIPP includes permanent disposal of TRU waste and not
simply "to demonstrate many technical and operational
principles." The FSAR should clearly state this.
The statement in this section that "the studies and
experiments using simulated wastes...are discussed in
numerous publications by Sandia National Laboratories and
other project participants" is wrong. We are not aware of
any published report that lists, describes, or discusses
experiments with simulated or real TRU wastes. Room J .- .
.<. :\. experiments to evaluate the corrosion effects of brine on the , c, .
,?, , . ' . ~ I
: . ; '1 55 gallon drums cannot be considered to be a nsimulated .> <: :.;, , . . ,
, . ,,. : , z , 9 : ' , ; ;# 8 . , wastesu experiment. The heater experiments were designed to
. ,'.,, . . ,. r:# . , . h-
, . . ,'*' y simulate the high level waste and not the TRU waste.
As we have stated before (EEG comments on Draft FSAR,
October 1988, Chapter 1, #3, P. ' 7 ) , the decision to use WIPP
for permanently disposing of the waste should be based on
demonstration of compliance with the EPA Standards 40 CFR
191, Subpart 8, and not "until sufficient operating and
scientific data have been accum~lated.~ The Standards do not
require operating experience.
4. Section 1.1.3, Design Capabilities, Page 1.1-12.
Please delete all references to experiments which are not
described and made a part of the FSAR.
5. Section 1.1.4, Schedule, Page 1.1-13. This section
should indicate when experimental data and other information
will be provided which support the need for the pilot plant
phase of WIPP, i.e., ernplacing CH-TRU waste at WIPP.
6. Section 1.3, Page 1.3-1. "The shipments are
surveyed for external contamination prior to their movement
11
into the WHB..." During the preoperational audit this was
done at the gate. What is planned? EEG believes the check
should not be done at the main entrance gate (see our
comments on the Phase I1 Preoperational Appraisal).
7. Section 1.3.1. It is noted that references to
backfilling the waste storage rooms that were in Amendment 9
of the SAR have been deleted from the FSAR. We know that the
Department has decided to emplace the experimental CH-TRU
waste without backfill to avoid crushing the drums during the
retrieval period. The FSAR should clearly state this
decision and the reason for it and should state that when the
waste is emplaced for disposal, a properly selected, tailored
backfill will be used to fill the space between the drums,
above the drums, and between the walls of the drums. The
FSAR should also state that only the amount of waste
expressly needed for conducting experiments to help in
performance assessment (to show compliance with the EPA
Standards 40 CFR 191, Subpart B) will be emplaced in a .- -
,$ i- temporary mode without backfill. [ , . . ,~.. i*,: 's;!:, .;i, ..
6. .>., " ,,4\ ' , . . .. 8. Section 1.5.2.1, Repository Plugging and Sealing ,, t ' , .:,;.i. ,
?( % . ' L V! . ,
" . . .' ..*' Studies, Page 1.5-4. It would be helpful to include in this
discussion an indication of when the final decision on plug
and seal design will be made. Such a decision is important to the Final Safety Analysis.
9. Section 1.5.3, Site and Design Validation Ac-
tivities, Page 1.5-9. The last paragraph should be revised
to delete the indication that there is no recognized function
for crushed-salt backfill. There is a well recognized value
for such backfill as was discussed in the paper by Chatur-
vedi, Channell, and Chapman (1988) published in the Waste
Management 1988 Conference Proceedings.
10. Section 1.5.3.1, Site Validation Program, Page 1.5-
11. In the discussion of Hydrologic Tests, it should be
pointed out that the piezometers in the ChSH shaft have not
performed well. The groundwater pressures measured in the
water-bearing zones of the Rustler were approximately the
same as pressures measured at levels where no groundwater is
supposed to exist. Also, please clarify the statement that
"pressure changes could be diagnostic of changing conditions
in the rock or deterioration of seal materials."
11. References for Section 1.5, Page 1.5-15. This
reference list includes several unpublished documents which
are not available for evaluation. Yet, the discussion in
Section 1.5 failed to recognize many of the published works
of EEG which are-quite relevant to the topics discussed.
CaAPTZR 1A
SuPlPary Safety Analysis
A. General Comments
This Summary Safety Analysis Chapter reflects items and
conclusions covered in somewhat greater detail elsewhere in
the FSAR. Our reaction to the presentation and conclusions
of the individual items is expressed elsewhere in comments on
the other chapters and is usually not repeated below.
1. This draft responded reasonably well to our October
1988 comments on Chapter 1A. Responses were made to 7 of our
11 comments; with- partial responses to two others. However,
all responses were not completely satisfactory.
.i I- ' -. I'
"2 " . , . ',
: $>% ', $, : .>I.'? f . '. Our first comment asked for a summary of the criteria ! ., !,;,%,. , b .. % used to determine that the facility could be operated safely.
This was partially answered by mentioning original site
design criteria. However, there are other safety-related EPA
regulations and W E Orders the facility will be required to
meet. What are they? Have you shown that you met them?
2 . This draft of the FSAR appropriately includes an
assessment of occupational and public exposures to the
hazardous waste component of TRU waste; yet, this is not
mentioned in Chapter 1A. It should be.
3. Section U . 1 . 1 . 2 , Wind, Page l A . l - 5 . The exhaust
filter building is mentioned elsewhere as a Design Class I1
Structure (not 111). Also, since it is Design Class 11, why
is it not designed for a 1 1 0 mi/hr wind?
4 . Table 1A.3-1, Page U . 3 - 3 . The title and column
heading in this table are confusing. What is being reported
is the average dose for six CH-TRU workers (four waste
handlers and two radiation control) and three RH-TRU workers
(two waste handlers and one radiation control). See Table
6.1-9.
5. Section lA.4.5, Waste Hoist Hydraulic Brake System,
Page 1A.4-5. The discussion on hopefully reducing the
estimated waste hoist brake system annual failure rate from
2-73-02 is not convincing. All that is expressed is a
feeling of faith that the total system failure rate will be
shown to be less than 1E-06 per year. See our comments under
Section 7.3.2.
6. Section 1A.4.5, HVAC Waste Handling and Exhaust
Filter Building, Page lA.4-6: The estimation that the
unavailability of the Exhaust Filter Building would be only
1.4E-04 per release event is unproven. It is understood that
this system did not perform properly during the March 1989
Preoperational Appraisal Audit. Furthermore, no allowance is
made for the failure of the CAMS to deliver a proper signal
in the event of a release. The ability of the CAMS to
perform with the required sensitivity in the underground
environment has yet to be proven. Also, EEG would appreciate
receiving information on the relay test circuits "being
considered." How can the problem be considered solved when
the specific correction is still being considered?
-2
Site Characteristics
A. General Comments
1. This Chapter has been greatly improved, and we were
gratified to note that many of the improvements were in
response to comments and recommendations of EEG.
B. Detailed Comments
1. Section 2.1.2.1.2, Potash Leases, Page 2.1-7. This
section does not provide adequate discussion of the potash
leases. Beoause there is indication that these private
leases may involve large sums of money, it is important that
they be resolved before the first shipment of vastes to WIPP.
Please add information on when a resolution is anticipated.
Also, add a description by section numbers of the 1600 acres
referred to.
2. Section 2.2.3.1, Fort Bliss/Biggs AAF, Texas, Page
2.2-2. We understand that there have been 14 flights per
year of UH-1H aircraft from the Biggs Air Force Base which
fly 500 ft. above the ground directly over the WIPP site.
These flights pose a threat to the safety of the WIPP
facility, and W E should take steps to insure they will not occur in the future.
cEAPJmt3
Principal Design Criteria
A. General Comments
1. Although Chapter 3 was extensively revised since
the earlier version of the FSAR, and there were many
editorial corrections, there was little substantive response
to EEG's previous comments and recommendations. We continue
to be opposed to the Design Classification definition of
Class I items, and believe that the definition should be
consistent with that contained in lo CF'R 60.
2 . EEG would appreciate the opportunity of reviewing
the design classification evaluations for all Design Class I1
items prepared in accordance with WIPP PROCEDURE WP-300,
CHAPTER 4. Please provide information on where this
documentation may be obtained.
3. Additional sections should be added to describe the
application of the Quality Codes 1, 2, and 3 that are
assigned to on-site work requests, to design documents, and
to purchase requisitions and for certain analytical or
laboratory services for Dasign Class I, 11, and IIIA facility
SSC's. Also, the Quality Codes should be described as they
apply to certain services associated with the design
validation, environmental monitoring, radiological monitoring
and geological programs. An additional section should also be added, either here or in Chapter 11 to describe the
implementation of the Quality Assurance Surveillance Program.
4 . The total expected inventory of CH-TRU and RH-TRU
wastes is never stated directly in Chapter 3 or elsewhere in
the FSAR. One can make estimates from various sources,
including: (1) Tables in Section 3.1: (2) Table 6.1-4: and
(3) statements about design volumes, operating lifetime, and
maximums permitted (for RH-TRU). But none of these calcula-
tions are as likely to lead to as good an estimate as that
determined by those in the W W who are most knowledgeable
about waste characteristics, generation rates, and treatment
and emplacement plans. Inventory data affect several aspects
of safety analysis: (1) estimates of the number of transpor-
tation and operational accidents; (2) probabilities of given
concentrations of radionuclides being involved in accidents;
and (3) the amounts of radionuclides available for release in
post closure scenarios. The FSAR should summarize as much
detail as is available on quantities, curies, and distribu-
tion in containers for both CH-TRU and RH-TRU. Possible
ranges and uncertainties should be discussed in all these
values.
B. Detailed Comments
1. Section 3.1.1.1, Container Configuration. Pages
3.1-3,30,31. The text on page 3.1-3 and the Tables 3.1-1 and
3.1-2 on pages 3.1-30,31 do not give a complete picture of
the CH-TRU containers since the TRUPACT Efficient Box
(Standard Waste Container) is not mentioned on any of these
pages. Also, several of the approved CIi-TRU vaste containers
(Table 3.1-2) cannot be shipped in TRUPACP I1 and it is
.. unclear how the vaste in these containers will be transported
to WIPP. These pages should be updated and adequately
describe the present situation.
2. Section 3.1.1.2.4, Thermal Power, Page 3.1-6. It
is unreasonable to assume an average thermal power of 60w for
RH-TRU if the average concentration is really similar to that
shown in Table 3.1-5. The watts from the Table 3.1-5
inventory would be about 0.6/caniSter. Furthermore, a
maximum canister subject to the constraints of 1000 PE-Ci, 23
Ci/l and 200 gm fissile material would produce only about 86
watts (assuming the same fission and activation product
mixture as in Table 3.1-5). A 1000 PE-Ci CH-TRU drum would
generate about 36 watts with no MAP or MFP radioactivity.
3. Table 3.1-5, Page 3.1-35. The average curies per
canister of RIi-TRU waste has been reduced from 1000 in the
previous draft FSAR to 37 (47 if daughter radionuclides are
included). This drastic change is not discussed in the FSAR.
The current Table is closer to the values reported in the
latest Integrated Data Base and in DOE/WIPP 88-005 but there
are a lot of internal inconsistencies within these documents.
It can only be concluded from the various reports that the
volume of Hi-TRU- coming to WIPP may be as little as 2500 m3
or it may be over 5000 m3. The number of curies may be from
less than 50,000 to over a million. There are other
- uncertainties; for example, some of the waste reported in
WE/WIPP 88-005 do not appear to be TRU (less than 100
nCi/gm) . "-.,,-- .-. , . . . ,'% , '
, p .., 2 The FSAR should explicitly discuss what is known as
f p. .. , . . . well as the uncertainties of the total RH-TRU inventory ', h? : ' :. '.. , . , , . ; expected to come to WIPP. This discussion should include
. . -. ..,.. _ estimated ranges of volumes and curies expected to come to
WIPP during its lifetime.
4. Table 3.3-2, CMS Vital Information Processing,
Pages 3.3-31 through 3.3-34. This table should be revised to
tabulate the CMS functions for each of the systems described
in Section 3.3.2 Air Handling, i.e., Section 3.3.2.1 Surface
Ventilation Systems for the Radioactive Materials Area,
Section 3.3.2.2 Surface Support Facilities Ventilation
System, Section 3.3.2.3 Subsurface Facilities Ventilation
System, Section 3.3.2.4 Interactions Between Air Handling
Systems. The present table appears to incorporate all of the - air handling functions in the one designation, HVAC, which
could be interpreted to be only the surface air systems with
air conditioning. The CMS functions for the Subsurface
Facilities Ventilation System, Section 3.3.2.3 are detailed
on pages 3.3-12 through 3.3-16 and should be tabulated in
Table 3.3-2.
On page 3.3-32, Table 3.3-2 should be revised to
tabulate the CMS functions for each of the four Shaft & Hoist
Systems identified in Table 3.1-8, i.e., Waste Hoist,
Construction h Salt Handling Hoist, Exhaust Shaft, and Air
Intake Shaft. For the Waste Hoist, the CMS functions should
indicate when TRU waste is being transported, when there is
hoist or shaft maintenance, shaft inspection, and. personnel
transportation.
5. Section 3.4, Decommissioning and Decontamination - Design Criteria, Pages 3.4-1 through 3.4-3. The Decommis-
sioning and Decontamination Design Criteria section should
discuss and reference the design criteria and programmatic
requirements of DOE Order 5820A, "Radioactive Waste Manage-
ment, Chapter V, Decommissioning of Radioactively Con-
taminated Facilities." The current reference to Chapter 12,
which references DOE 5820.2A1 is considered to be inadequate for the purposes of the FSAR since neither Section 3.4 nor
Chapter 12 address each of the major design and programmatic
requirements (5.1 through 5.e) of DOE 5820.2A.
The reference to the ALARA program should reference DOE
Order 5480.11, "Radiation Protection for Occupational
Workers," which references (paragraph 9.a) PNL-6577, "Health
Physics Manual of Practices for Reducing Radiation Exposure
to Levels that are As Low as Reasonably Achievable ( A m ) . "
aAPTER4
Plant Design
A. General Comments
1. This chapter has been expanded to include sig-
nificantly improved descriptions of ventilation, fire
protection, electrical and water distribution. This
information was quite helpful to EEG1s review.
2. Inspections and testing &f important equipment and
mechanical systems is discussed throughout this chapter, but
only occasionally is the frequency given for such inspections
and tests. This additional information should be added for
all of the important systems, or a reference cited where such
information is specified.
B. Detailed Comments
1. Figure 4.1-2, WIPP Surface Structures, Page 4.1-8.
Buildings 364 and 365 are not identified on the Explanation
page - 2. Section 4.2.1.1, Inventory and Preparation Area,
Page 4.2-4. The description of this area refers to a waste
surge storage area, a battery recharge area, and office space
for waste handling personnel. These areas should be
identified in Figure 4.2-1 or reference made to other figures
where the areas are identified.
3. Section 4.2.1.2, RH Waste Handling Area, Page 4.2-6
to 4.2-10. Figures 4.2-1 and 4.2-3 should be changed to
label and identify the cask preparation area, the cask
maintenance station, and the cask transfer cell.
4. Section 4.2.1.5, Fire Protection, Page 4.2-14. The
last paragraph of this section discusses solidification of
contaminated liquid wastes. As previously requested, the
FSAR should provide further details of this processing - either here or elsewhere in the FSAR. It should address
radiation protection and include the processing procedure,
where such processing will take place, and how separation
will be maintained of contaminated RH-waste water from
CH-waste water.
5. Section 4.2.1.6, Effluent Monitoring System, Page
4.2-15. The air that exhausts the WHB is filtered by a
prefilter and two HEPA filters, not "... 3 multiple stages of HEPA filters ..."
6. Section 4.2.2.2, Construction and Salt Handling
Shaft Headframe and Hoist House, Page 4.2-17. Please provide
further clarification of the "placement of the emergency - escape hoist over the CLSH shaft." Is the emergency escape
hoist not part of the existing/in-place mucking hoist, or
will there be a requirement to change cages prior to
evacuating the mine through the C&SH shaft?
7. Figure 4.2-7, Support Building, Page 4.2-34. Rooms
124 and 125 are mislabeled. The High- and Low-Level Counting
Labs are not located here. The walls that remodeled rooms
107 and 139 into offices are not shown.
8. Figure 4.2-8, Support Building, Page 4.2-35. The
floor plan for rooms 250 through 253 are incorrect.
9. Section 4.3.2.1.2, Electrical Utility Services,
Page 4.3-9. The description of the Electrical Utility
Services should be revised to state that one diesel generator
can be remotely started and brought on line from the Central -
- Monitoring Room, as stated in Section 4.4.2.1.2. As shown on the legend to Table 4.4-6, DG No. 25P-E504 is able to be
synchronized and brought on line automatically.
10. Section 4.3.2.1.3, Subsurface Structural Features,
Page 4.3-10. The third paragraph on this page suggests that
vapors from diesel fuel constitute the principal risk of an
underground explosion. Of possible greater risk is an
explosion from hydrogen formation around the battery recharge
areas. Battery recharge in the WHB is shown in Figure 4.4-2
and discussed in Section 4.4.3.1.2, but no mention is made of
subsurface battery charging. This should be addressed here
and in other sections of the FSAR. This deficiency was
mentioned in previous EEG comments. Also, it was previously
brought out-that the wastes may contain small amounts of
pyrophoric materials, or materials which may produce
explosive mixtures. Since backfill will not be used during
- the first five years, some consideration'of this potential problem should be discussed.
11. Section 4.3.2.3, RH-TRU Waste Storage Area, Page
4.3-13. EEG requests that up-to-date reference design
documentation be cited on pages 4.3-13 and 4.3-15, and that
EEG be supplied with drawings of the current design of the
shield plug and documentation that supports the 3 - 5 mrem/hour statement.
"
12. Section 4.4.1, Ventilation Systems, Page 4.4-1. The
last paragraph on this page refers to UP 04-1, "Facility
Operation Manual." This is an important document which EEG
needs to complete review of the FSAR. As of this date, it
has not been made available.
13. Section 4.4.1.2.1, CMR and Instrument Shop, Page
- 4.4-10. The last paragraph should be revised to reflect the
fact that the CAMS located at stations B and C do not monitor .-..~
ambient air, but monitor only HEPA-filtered qqcleanqq air and
therefore do not necessarily represent ambient air.
14. Section 4.4.1.3.2, System Description, Page 4.4-18,
4.4-19, 4.4-20. On page 4.4-18, the statement, "Alarm of any
two CAMS can activate HEPA filtrationm is misleading.
According to J.P. Harvill, at the 28th Quarterly Meeting
between W E and EEG, an alarm by any two underground alpha or
any two underground beta/gamma CAMS will initiate HEPA
filtration.
On page 4.4-19, please clarify how the reversal of air
flow will impact the emergency traffic flow underground
pursuant to the escape markings.
The last two paragraphs on page 4.4-20 refer to
nperiodicq8 leak and operational tests. Please indicate the - frequency of testing, and appropriate action levels.
15. Section 4.4.2.1.5, Backup Loads, Page 4.4-25. This
discussion of the backup loads should be revised on page 4.4-
25 and in Table 4.4-8 on page 4.4-126. The electrical power
needs of the Exhaust Fans, items 28 and 29 on the Legend to
Table 4.4-5, page 4.4-105, should be identified rather than
the backup electrical load for ventilation for the Air Intake
Shaft Hoist on page 4.4-25 and the Air Intake Shaft Hoist
Fans on Table 4.4-8. Also, the discussion of the minimum
load for backup should be consistent with the above revision.
16. Section 4.4.3.1.2, Fire Characteristics, page 4.4-
32. The discussion of Svontaneous ianition includes mention
of hydrogen formation and venting through a separate exhaust
system. It is presumed that this "separate exhaustn venting
of hydrogen applies only to the battery recharge area of the -
WHB. Please also indicate what safety design features are
available in the subsurface battery recharge areas.
17. Section 4.4.6, Radioactive Waste Systems, Pages
4.4-49 and 4.4-50. This section refers to section 5.4 for a
description of decontamination procedures, process operation,
and radwaste properties. Section 5.4 anticipates radioactive
waste will be generated above and below ground at WIPP,
however no mention is made of how water mixed with waste and salt in the underground will be collected, assayed or
solidified, nor how the underground tunnels will be decon-
taminated.
The last paragraph of this section on page 4.4-50 refers
to the "Waste Handling Operations ManualIn WP 05-1. The EEG
has not been provided a copy of this manual which is
necessary to complete our review of the FSAR. The section
also refers to the FMEA, Table 4.4-13. This table does not
appear to address the waste water from the RH area, which is
contaminated and may have to be treated separately from the
CH-liquid waste. This problem should be addressed.
18. Section 4.4.6.2, Liquid Radioactive Wastes, Page . ,. 4.4-52. This section describes a trench system which holds
, , .. fire water pending sampling and analysis for radioactive . . . ' , ,
i i. :..I . r ,\. ~, , # :
, : . . i , . , . contamination. If contamination is confirmed, then the
I ...I ~. , ' ' .:.* :.: ; . , , ; , I .,; a,, : contaminated water is manually transferred to a collection
'. . ."..*.... . ,-
tank. This section does not provide details of radiation
protection for workers, and procedures needed to collect, mix
and measure the activity of the supernatant or precipitate in the holding tank and sump.
19. Section 4.4.9.4, Air Filtering Equipment, Page 4.4-
63. Please provide more information on the criteria for
filter changeout - the radiation level or pressure drop. 25
This information should either be provided here or a
reference to where such information is available should be
provided. Also, it is noted that there is no airlock
included in the present design of the Exhaust Filter Building
HEPA filter plenum. Further details of the changeout
procedure are needed.
20. Section 4.4.9.4, Air Filtering Equipment, Page 4.4-
64. It seems likely that the negative pressure of the filter
plenum in the Waste Handling Building would collapse the bag
for the used filters.
21. Section 4.4.10.2.2, Exhaust Filter Building, Pages
4.4-69, 4.4-70. This section indicates that the compressed
air requirements are met by two compressors. It was EEG's
understanding that compressed air to the EFB is provided by a
buried pipe to the Compressor building.
I
22. Table 4.4-1, FMEA for the WHB HVAC System, Page
4.4-79. Item 6 on this page describes a Failure Mode as
"Permissive to supply air handling unit fails." Please
clarify this failure mode.
CBAPTW5
Process Description
A. Detailed Comments
1. Section 5.1.1.2, Inventory/Preparation Area, Page
5.1-4. Further details are needed here or elsevhere in the
FSAR on the procedures for removal and assay of the HEPA ~ " .. - filters to avoid potential contamination.
2. Table 5.1-3, CH-TRU Waste Handling Failure Mode and
Effects Analysis. This table fails to consider the potential
for fire or explosion resulting from hydrogen gas around the
battery recharge area in the subsurface. According to
Figure 4.3-5, there are five battery recharging stations in
the subsurface.
3. Table 5.1-3, CH-TRU Waste Handling Failure Mode and
Effects Analysis, Pages 5.1-16, 5.1-19. On page 5.1-16,
accident 13 refers to use of breathing air masks for fire in
a site-generated rad waste room. Only SCBA would be approved
for use during a fire. Please clarify.
On page 5.1-19, Item A1 for Accident 19 is not a safety
feature as stated.
4. Table5.2-1, RH-TRU Waste Handling Failure Mode and
Effects Analysis. This table also fails to consider fire or
explosion in the underground battery recharge areas.
See previous comment.
5. Section 5.4.2, Solid Radwaste System, Page 5.4-2.
This section states that all solid radwaste is anticipated to
be CH, however, if contamination is found or occurs in the
hot lab or RH-canisters, the resultant cleanup could produce - RH-waste. This should be reevaluated and a procedure
developed to handle RH-TRU generated wastes.
6. Table 5.5-1, Waste Package Information, Page 5.5-7.
This table omits several items which are required pursuant to
WIPP/DOE -157, Rev. 2. For example, the Shipment Certifica-
tion Date, the name of the official who certified the TRUPACT
payload, the organic materials volume present, the thermal
power (if the amount exceeds the specified limit), the Pu-23.9
Gram Equivalent, the Waste Package Certification Date, and
name of the certifying official are either required or
conditionally required.
- 6
Environmental, Safety and Health Protection
A. General Comments
1. Appendix 6A, which is necessary for the review of
the radionuciide concentration and dose calculations in this
chapter, was not received until April 26, 1989, after most
comments for the chapter were completed. Our comments on the
appendix are summarized at the end of the chapter. We 5, consider the appendix to be seriously flawed. f
3 x 4) 9 . L1
/' 2. There appears to have been a failure to address in " -
this chapter the-changes brought about by the introduction of
the new ventilation shaft and the new fans. Also, there is a
need for more careful consideration of the placement of
monitoring equipment in the Waste Handling Building. There
should be a systematic ventilation and contaminant migration
study with smokes and tracer gases to arrive at more
realistic decisions on placement.
3 . The Draft FSAR makes only two brief references to
the requirements of 40 CFR 191 Part A and does not explicitly
say how compliance will be shown. Neither does the Draft
FSAR compare expected doses estimated in this chapter with
Part A. The Second Modification of the C&C Agreement
requires that the FSAR document comply with Subpart A.
Therefore, this documentation of compliance must be included
in the FSAR.
4. The Draft FSAR does not fully cover the disposal
phase of the project because compliance with 40 CFR 191,
Part 8, has not been shown, and final decisions have not been
made on waste treatment, backfill, and emplacement details.
The need for a supplement to the FSAR has been recognized in -.
the Draft FSAR and is included elsewhere in our comments.
However, this Draft FSAR does not address any of the
operational procedures that will take place in the proposed
experimental phase of WIPP. There are important differences
in the waste form that would be used for proposed bin
experiments and in underground handling procedures for both
experimental and proposed operational demonstration wastes.
In addition, the movement and/or backfilling of wastes
emplaced during the experimental phase into the final
disposal mode must be evaluated.' There are possibilities of
increased radiation exposure and perhaps mine safety when
working in rooms that will have been open for six to eight
years. In addition, treatment of waste containers on-site is
a possibility. Also, some emplacement rates which have been
proposed during the experimental phase could lead to a
three-panel operation during the first few years when
experimental phase waste is being finally emplaced and new - waste is being brought in for the disposal phase. The
adequacy of the ventilation system to allow waste handling
operations in three panels needs to be evaluated.
B. Detailed Comments
1. Section 6.1.1.2, Design Consideration, Page 6.1-2.
In view of recent moves to super-compact waste and heavily
load boxes, and the possibility that the existing inventory
of boxes may be repacked into the smaller TRUPACT-I1 standard
waste containers, it may be prudent to reevaluate the I*.
assumption that radiation fields from boxes will be smaller
than from a 55 gallon drum.
2. Section 6.1.2.1, Direct Radiation Sources, Page
6.1-5. The third paragraph uses %r/hm as allowed neutron
dose rate. This should be *mrem/h.* Also, please provide -.
the basis for ignoring the neutron contribution to total dose
rate, particularly from high alphacontent (heat source Pu-
238 and enhanced Am-241) wastes.
3. Section 6.1.3.1, Plant Arrangement Designs for
Keeping Exposures ALARA, Pages 6.1-8, 6.1-9. On page 6.1-8,
there is a need for further information on the contamination
check points. For example, describe the equipment to be
used, the procedure for survey of personnel and control of
potential contamination, and the procedures and facilities
for handling contaminated personnel when or if found. On
page 6.1-9, please include information which supports the
assumption that pressure differential values created by the
ventilation system correspond correctly and appropriately to
the identified c~ntamination 88zones18 in the WHB such that
contamination spread between zones will be controlled.
- 4. Section 6.1.3.3, Radiological Control Zoning and
Access Control, Page 6.1-12. More detail is needed here on
how personnel are surveyed for contamination at the control
points, equipment used, what action is taken when con-
taminated workers are found (where decontamination occurs),
etc.
i #
5. Section 6.1.3.4, Radiation Shielding, Page 6.1-17
to 6.1-18. It is very difficult to review input parameters r . %
for the operation of the QAD-P5A computer code from figures
6.1-6 through 6.1-8. A listing of the structural and
configurational, as well as source term magnitude/location
for this analysis, and those involving the execution of two
other shielding codes (63 and ANISN) is required to make an
analysis of this activity with any degree of confidence.
6. Section 6.1.3.5, Ventilation, Page 6.1-20. Please
- clarify and revise the grammatical errors in the first
paragraph on this page.
7. Section 6.1.3.6, Radiation Monitoring Equipment,
Pages 6.1-24,25. The discussion on page 6.1-24 concerning
the placement of alpha CAMs and FASs is misleading and
implies a lack of understanding of the purpose of these
sampling systems in meeting the monitoring requirements of
W E Order 5480.11:
(a) CAMs are not designed to provide indications
of concentrations of airborne radioactivity. Their role is
to provide alarm in the event of accidental releases:
(b) The decision to utilize a CAM, according to
W E Orders, is not based on whether personnel occupancy is
nlowll or not: and
(c) FA5 sampling is not an adjunct to CAM
monitoring in cases of "low occupancy." FASs havetheir own
proper function in monitoring and control of worker exposure.
This should be clearly described here.
The second paragraph on page 6.1-25 states that each
monitoring system is set to alarm within "acceptable levels
of the limits in W E 5480.1B, Chapter XI." Since DOE 5480.1B
does not address alarm levels, please provide an indication
of the criteria which will be used to establish these levels.
For example, perhaps the level will be set at some designated
fraction of the maximum permissible exposure range. It would
be desirable to specify that fraction either in the FSAR or
the Radiation Safety Manual, so that it could be verified and
''.;Fld not become an arbitrary value.
Also, the discussion of airborne radioactivity monitor-
ing assumes that the monitors are "designed to operate in the
expected environmental conditions." Based on recent reviews
of the CAMS for both collecting and detecting transuranics in
a radon daughter and salt loading environment, this assump- -
tion is unproven. Also the calibration of the CAMs, while
traceable to NBS and which provide instrument checking, do
not calibrate for the actual environmental conditions.
Proper operation of the CAMs is vital to the protection of
workers at WIPP and to the warning of environmental releases.
EEG's concerns about the ability of these instruments to
detect radionuclide concentrations with the required
sensitivity (particularly in the repository where the
greatest amount of interference from salt loading and radon
daughters are encountered) have been well documented
elsewhere. It is mandatory that further studies be carried
out to insure that an adequate monitoring system will be in
place before wastes arrive.
8. Section 6.1.4.2, Normal Operation, Page 6.1-27. In
the description of the input for annual exposure during
normal CH-TRU and RH-TRU waste handling operations, dose
rates of 14 mrem/h, 5 mrem/h, and 2 mrem/h four inches from
the surface are given for CH-drum, CH-box, and RH-cask,
respectively. Please provide the basis for these values.
Also, please clarify whether the surface level on a drum is
from an individual drum or a 7-Pack. ".I---..
% *q i$%,
' 3 , ? :: , .:< a , ;. i 9. Section 6.1.5.3, Radiation Protection Instrumenta- : ?; $; ::. ..r ,; ! .. : . . : . tion, Page 6.1-34, 6.1-38.
, . . . , , . . ,
On page 6.1-34, it is stated that samples for both the
low level laboratory and the high-level laboratory are
prepared in the sample preparation room. It is unacceptable
to prepare a sample for high-level counting in the same room
as samples for low-level counting since cross contamination
of the low-level facilities will likely occur.
Please provide a reference where further details for the
calibration procedures may be found.
The first paragraph on page 6.1-38 states that bioassay
services are available on a contract basis. The implication
is that bioassays would be performed only following evidence
of a contamination incident. The WIPP "Radiation Safety
Manual," WP 12-5, Revision 1, indicates that preassignment
baseline assays and annual bioassays will be routinely
carried out. This would suggest that on-site capability for
certain types of bioassays (urine, fecal and chest counting)
should be available, and the annual bioassays for radiation
workers scheduled to minimize assay work loads. This section
should be revised to clarify that routiw and incident assays
will be carried out pursuant to the "Radiation Safety
Manual." The discussion on page 6.1-41 clarifies this
ambiguity tu some degree, but has a typographical error in
the sixth paragraph where Nroutingn should be Nroutine.ll
10. Section 6.1.5.5, Radiological Control Facilities, .--. Page 6.1-43, 6.1-44. Please provide more information on the
facilities and methods for personnel decontamination, and for
detecting such contamination. With respect to personnel
decontamination, reference is made to the transport of
patients to "a hospital, which has agreed to handle injuries
involving radioactive materials.*# Has a hospital(s) been
identified for this purpose? If so, please identify it in
this document.
On page 6.1-45, under equipment for the dosimetry
laboratory, a statement indicates that compressed nitrogen
gas is used for heating. Please clarify this statement.
11. Section 6.1.6.1, sources of Potential Release, Page
6.1-48. The
in assessing
provides the
first paragraph states that the assumptions used
releases are shown in Table 6.1-5. This table
estimated results only, but does not indicate -
the basic assumptions used to obtain these values. The basic
assumptions are needed in order to properly evaluate these
data. Neither Table 6.1-4 or Appendix 6A provide sufficient
information to arrive at the data shown in Table 6.1-5. Also
the first sentence in this paragraph should be revised to
state that @@the design of WIPP recognizes that very small
amounts of radioactivity will be released.@@ To state that it
"may1' be released implies that no radioactivity may be
released, which is not possible.
12. Section 6.1.6.2, Dose Calculation Models, Page 6.1-
51. The use of the mean reciprocal wind speed in the
atmospheric dispersion equation instead of the mean wind
speed biases the equation toward the lower wind speeds. The
resulting deposition would be higher than if the numerical
average of the wind speeds were used in the equation. Please
clarify how the average of the reciprocal wind speeds can
save computer time, or other reason why these values were
used.
Also on the following page (page 6.1-52), in the
determination of the effective stack height for air dis-
charges, credit is taken for the effluent air velocity in the
vertical direction. It is our understanding that these stacks are not arranged to release air in the vertical
t direction, nor will they emit air equally well in all
directions. Therefore, the angle of discharge, and the
effects of shrouding to effect releases in one direction from
the stacks, should be taken into consideration when utilizing
the Rupp equation to determine effective stack height. Lower
effective stack heights and greater momentum of air in the
horizontal direction (and possibly turbulence when the
airstream must change direction) after exiting would be
expected. Some consideration should be given toward
determination of the exiting plume as a function of wind
direction and velocity since the dry deposition rate 1s -. affected by wind speed in the horizontal direction as well as
by radionuclide concentration.
On page 6.1-53, the use of a constant precipitation
factor for the determination of radionuclide deposition is
questionable for several reasons. The first reason is stated
in the document: precipitation occurs in discrete events of
varying magnitude throughout the year. At the WIPP, there
are about three to nine events/month on the average (Climates
of the United States, U.S. Dept. of Commerce, Washington,
D.C. 1973) depending on the month of the year. However, most
of the precipitation, and the greatest number of
precipitation events, occur during the growing season.
Hence, most'of the deposition of this type occurs during
livestock foraging periods and does not occur equally
throughout the year. This period also marks the greatest
surface contamination of forage plants which are consumed by - livestock. During the spring when biomass densities are
limited, livestock must forage over large areas to obtain
their food requirements when compared to later in the growing
season. During the growing season the deposition is affected
by the leaf area index of the plant on the one hand, and
lower grazing areas/animal unit as well as tissue dilution of
the areal radionuclide concentration on the plants. Thus a
complex set of processes involving both precipitation events
and biomass densities affect livestock intake of
radionuclides deposited on plant surfaces. A simplistic
approach in using a constant scavenging coefficient is ,#--.--.
/ $&,, probably not conservative, and should be tested with a model
f-.::. $:..'; , 1; .$. ' ') that can take these factors into consideration in evaluating ,.::,.w.
i , . .- - : . ." this dose pathway to man to give it validity.
,.,. . - This document does not mention, and presumably does not
include, resuspension events which can contaminate plant
surfaces via additions to the air concentration resulting
from stack emissions. They are mentioned only with respect
to inhalation hazards to man. Curiously, it does mention
washoff of contamination from these surfaces in estimating
pathways. The document is totally silent on soil saltation-
creep (erosion) events which contaminate forage and crops up
to about one meter above the ground, and also rainsplash
momentum which does the same thing. In a region where
radionuclides accumulate on the soil surface as a result of
deposition from a plume, these processes become very
important to consider as pathways of radionuclides to man.
It is well known that insoluble plutonium compounds are not
readily eluviated from the soil surface. Hence, this pathway
is ever present, particularly in an arid environment, and
should be t-aken into account in a risk model for this
purpose.
Although dose calculation equations from reference '(39)
were used, it is not clear what specific equations were used,
and what differences in parameter input assumptions were
made. This should be clarified. Also the meaning of
"exponential transferfrom one segment to anotherw is not
clear. Presumably this refers to the four-segment catenary
model of the GI tract developed by Eve. This should be stated. Generally, the integrated form of transport
processes can be expressed as exponential equations with
constant or time variant coefficients, but the actual
transport processes are not of this type.
13. Section 6.1.6.3, Dose Calculation Parameters and
Discussion of Results, Page 6.1-55. The assumed value for
the deposition velocity at the WIPP site of 0.18 cm/sec was
not found in reference (38), however, a value of 0.68 cm/sec
was found for typical meteorological conditions at Oak Ridge,
Tennessee. This value for the WIPP site appears to be low by
3 7
a factor of about two. A justification and an exact
reference for this parameter should be added. It is
generally assumed that there is a one order of magnitude
difference between wet and dry deposition amounts (wet = 10 x
dry) . It is not clear whether the 0.28 kg/m2 and 1.9 kg/m2
values on page 6.1-56 are wet or dry weight units. The value
of grassland biomass density appears to be greater than
expected by a factor of two for arid sites if wet weight
units are assumed. Please clarify whether the reference
cited specifically addresses arid sites. The value used for
forage consumption for cattle is 15.6 kg/day dry weight. The
value recommended by NRC is 12.5 kg/day. Why was the latter
value not used? Also, it is not clear why it was assumed
that the entire beef herd was consumed on an annual basis, or
why a specific fractional part of the herd was assumed to be
consumed per day if one assumes constant biomass density and - an adequate number of beef cows are available to feed the
individual(s). Because beef cows are not slaughtered until
they have reached a certain stage of maturity (discrete
event), and a human's beef consumption is continuous, a lag
period accounting for maturation (seasonal), and another
between slaughter and consumption (NRC recommends 20 days,
although it is not important for transuranics) should be
implemented. Please indicate how the seasonal factor was
addressed in these analyses. If mean annual concentrations
in beef tissue are being utilized each day to account for the
maturation period, the validity of this assumption should be
documented. Please also indicate the extent to which water
ingestion was considered in these analyses.
14. Section 6.1.6'.4, Effluent and Environmental
Monitoring and Exposure Pathways, Page 6.1-57. The first
paragraph of this section states that the nonradiological -.
monitoring is discussed in Chapter 3. We could find no such
discussion in Chapter 3. There was a reference to handling
non-radioactive hazardous materials, namely the "Operations
Program Plan,It WE/WIPP 85-001, Rev. 3, July 1988. This
document has not been provided to EEG.
On page 6.1-62, it is not proven that either of the
alpha CAMS can correctly measure the release of TRU from
underground in the presence of salt loading of the filters.
In the case of the beta CAM, the correction for radon progeny
beta emitters is not discussed, and how the salt loading on
the monitor will affect the gamma correction on the opposed
t8nonloadedn detector. What are the lower limits of detect-
ability for the WIPP radionuclides in environmental media for
these detector systems?
On page 6.1-64, first sentence. This sentence should
read: *'The filters obtained from this FAS will be collected
and analyzed by the New Mexico Environmental Evaluation
Group, for independent verification of releases from the
facility."
On page 6.1-64, please provide more information on the
minimum detectable concentrations (MDC) which determine
quantitatively the meaning of the phrase "significant
release." Also indicate what total release and/or release
rate would correspond to these levels based on calculated X/Q
and deposition velocities for the predominant plume direc-
tion. It should be noted that once the HEPA filters are
activated in the EFB, Station B must remain in operation
continuously from that time on, because it is the sole point
at which this effluent can be monitored after it passes
through the HEPA filtration system.
15. Section 6.1.7.1.1, Overall Approach, Page 6.1-70.
Under item (4) it is stated that a 50-year effective -.
committed dose equivalent is used. This concept is commonly
used for radionuclides, but it should be made clear how the
methodology would be applied for hazardous wastes.
On page 6.1-72, the first paragraph includes a statement
that risks for hazardous wastes are being overestimated by
one to three orders of magnitude by the use of conservative
assumptions. The basis for this statement should be added or
the statement deleted.
16. Section 6.1.7.1.2, Assumptions and Considerations
of Uncertainty, Pages 6.1-71, and 6.1-72. The first sentence
of this section indicates that conservative estimates are
provided in-Table 6.1-16. Please provide a justification for
this statement. EEG agrees with the first sentence of the
second paragraph on page 6.1-72 in that this statement
recognizes the uncertainties in the chemical data base.
Also, the assumption of 24 hours-a-day, 365 days-a-year
occupancy may lead to a conservative factor of two or three,
but not a factor of 10 to 1000.
17. Section 6.1.7.2.1, Migration Pathways, Page 6.1-73.
We agree that for normal operations the volatile organic
gases would be the predominant releases, however, it is not
obvious that 100% of the lead would be in a monolithic form.
Since there is so much lead in the waste, the mobility of
only a few percent of it could be significant. A contrary
assumption should be justified.
Also there is no indication that surface contamination
has been considered. Data is needed on surface contamination
so that exposure can be evaluated.
This section concludes that the ingestion pathway was
not evaluated. This conclusion is based on the assumption - that the chemicals are relatively insoluble and tend to break
down in the atmosphere and through biodegradation. Please
provide data or references which support such assumptions.
18. Section 6.1.7.2.2, Characteristics of Potential
Hazardous Contaminants, Page 6.1-74. It may not be valid to
consider only those hazardous wastes present to the extent of
1% or more. It is possible that a highly toxic component
present in amounts less than 1% by weight may be of greater
significance in toxicity evaluations. For example, cadmium
has a cancer potency factor about 425 times as great as
methylene chloride and is listed in Table 6.1-17 as having an
average concentration about .006 of that of methylene
chloride. Therefore, perhaps cadmium should have been
considered. Was the data base examined to determine the
presence in low concentrations of other highly toxic waste?
A
The statement in the second paragraph on this page that
the method of calculation leads to a "worst case1* scenario is
correct for some of the waste forms but not for those waste
forms in which the concentration of a hazardous constituent
may be greater than the ncalculated" average. Also, the
concentrations of hazardous waste from other generators may
substantially exceed that from RFP/INEL. Therefore, more
convincing evidence is needed to support the claim that this a
is a conservative estimate.
At the bottom of page 6.1-74, it appears to be assumed
that lead is not one of the mobile constituents. With the
very high quantities of lead in some of the drums, this may
not be valid. Evidence should be provided to support such an
assumption. Furthermore, lead can be assumed to be control-
ling for the other heavy metals only if the mobile fraction
and/or toxicity of these other metals is substantially less
than for lead.
On page 6.1-75, a statement that volatilization of
liquid organics need not be considered because the WAC does
not allow liquids is inaccurate. Actually, the WAC does
allow small amounts of liquid residues.
As previously indicated, the statement at the top of
page 6.1-77 that using an average concentration represents a
worst case assumption is not valid for all waste containers.
19. Section 6.1.7.2.3, Exposure Modeling, Page 6.1-78.
Using the "nearest residence," the Mills Ranch, as the
location of the maximum exposed individual from routine
releases is'probably not conservative. The X / Q cmcentra-
tions at Crawford Ranch (5 miles NNW-NW in the prevailing
downwind direction) are about 502 greater than at Mills Ranch
(Table H-49, page H-93 of FEIS). For accidents, Mills Ranch .-
has one hour X / Q values about 108 higher than Crawford Ranch.
In the second paragraph on page 6.1-78, there appears to
be a typographical error in the first sentence. The volatile
releases should be "fromn the waste handling building and
underground area rather than "to." Also please address the
potential for adsorption of the VOC9s to particulates. In
which case, why should not the particulate form of hazardous
materials be considered? Please also clarify whether the
mixing heights utilized for VOC1s are the same as those used
for the dispersion of radioactive material. Do VOC's have a
higher effective mixing layer because of diffusion?
20. Section 6.1.7.3, Routine Releases and Exposures for
Hazardous Chemicals, Page 6.1-80. This section states that
"before opening the TRUPACT-11, samples will be taken from
the sample port to detect any accumulation of hazardous
chemicals." This monitoring procedure is not described in
Chapter 5. It should be indicated whether such analyses of
hazardous chemicals will be a routine procedure, and, if so,
what methods of analysis will be used. If these materials
will be routinely analyzed in the gases of all TRUPACP-I18s,
the procedure should be addressed in Chapter 5. The average
flow rates in Table 6.1-19 for 14 drums sealed in TRUPACT-I1
for 100 hours would lead to air concentrations in the
TRUPACT-I1 cavity that exceed threshold limit values for
carbon tetrachloride and are about 15% of the TLV for 1,1,1
trichloroethane. Therefore, concentrations inside the
TRUPACT-I1 may be significant and routine sampling should be
required.
21. Section 6.1.7.3.1, Routine Releases, Page 6.1-80.
The statement is made here that *'backfilling is expected to
effectively reduce exposure to VOCs to negligible levels.'*
This would be true only after a storage chamber has been
sealed, but not necessarily during the filling of a chamber
with waste.
Also on this page, there appears to be an error in the
assumed air velocity in a storage room. Since the empty room
has a cross-sectional area of about 40 m2, the velocity of 3 m/s yields a ventilation in the storage room of 120 m3/s or
254,000 ft3/min, which is 60% of the entire repository
ventilation air. Also using the data in Table 6.1-22, one
can calculate the air flow to be in the vicinity of 116 to . r t
' 123 m3/s. The WPO ventilation drawings indicate that the . 4
entire flow in a waste panel will be 122,000 ft3/min. Thus
khe air flow assumption here appears to be high by at least
- ' 108% and the calculated air concentrations in Table 6.1-22
should be more than doubled. (~eth~lene chloride concentra-
tion appears high by a factor of 10.)
22. Section 6.1.7.4, Health Risks and Ecological
Consequences of Chemical Releases, Page 6.1-83. Please
clarify this paragraph and indicate whether the acceptable
excess cancer risks (1 in 10,000 occupational, etc.) are per
year or per lifetime.
The assumption of 65 to 70 individuals exposed does not
appear consistent with information in Tables 6.1-8, and 6.1-
lo. Please clarify. Also the significance of the last
sentence on this page is not clear.
Concerning the discussion at the top of page 6.1-84, the
use of human risk standards when considering exposure risks
to animals may be conservative for certain species, and/or
hazardous materials, but not valid for others. For example,
herbivores do not wash forage prior to consumption and are
more likely to inhale resuspended contamination for longer
periods of time in a contaminated area. Therefore, it may be A
desirable to carry out additional environmental studies of
the ecological system to further support these assumptions.
On page 6.1-85, it was possible for EEG to derive the
risk factor for carbon tetrachloride as stated in the second
paragraph by using a cancer potency factor of 0.13 (mg/kg/d) 'l, and by assuming the values in Table 6.1-22 are
in mg/kg/d. However it was not possible for us to arrive at
the risk factor for methylene chloride. Please clarify these
derivations. Also, the statement that 2.73-06 is "at least
two orders of magnitude less than 1E-04," is not correct.
23. Tables 6.1-16 and 6.1-17, Pages 6.1-108 and 6.1-
109. By using the values in Table 3.1-6, it was possible to
derive the values shown in Tables 6.1-16 and 6.1-17. It is
recommended that these two tables reference Table 3.1-6.
24. Table 6.1-20, Page 6.1-112. Except for the values A of Freon $13, it is not possible to correlate the values in
this table with the respective emanation rates shown in Table
6.1-19. Please provide the methodology for deriving these
values or information to indicate why the respective routine
releases should not be the product of the number of drums
times the emanation rate per drum.
25. Table 6.1-21, Page 6.1-113. The inhalation values
given in this Table cannot be obtained from data in other
tables when using normal inhalation rates of 12 m3/d for
- ... occupational exposure and 22 m3/d for the public. Possibly ,* -
f' :, - these tables should have been labeled as mg/kg body weight 1 f r t a h e r day. , " " i
t
26. Table 6.1-22, Page 6.1-114. In evaluating this
Table, and as indicated in comment 21 on this Chapter, the
air flows appear to be high by a factor of at least two. 6 Also if the mg/day values in Table 6.1-21 should be mg/kg/d
and are multiplied by 70 for a 70 kg individual, and using an
AIC value of 6.30 mg/Kg/day the risk would be 5.1E-04 for
1,1,1 trichloroethane. The risk value for carbon tetra-
chloride was verified but we did not agree with the methylene
chloride value when using a cancer potency factor of .0143
(mg/kg/d)-l. Also, isn't the cancer risk the excess lifetime
risk from a 25-year exposure at WIPP, rather an excess annual
risk as stated in the footnote?
27. Table 6.1-23, Page 6.1-115. No information is
given on the EPA ISG Dispersion Model used to calculate off-
site air concentrations. The values reported in Table 6.1-23
could be obtained by using an effective X/Q factor of about
5E-08 s/m3. The X/Q factor used in the FEIS (Table H-49) is
about 6E-07. Even the equivalent X/Q factor used to
calculate individual radiation doses in this chapter (in
Table 6.1-13 from releases in Table 6.1-12) is over six times - greater (about 3.33-07 s/m3). It was not possible to check
the values for carcinogens. Also, as mentioned in comment 19
on this chapter, it appears that Crawford Ranch would have
higher concentrations than at Mills Ranch.
The risk values were verified for all hazardous
chemicals except methylene chloride by assuming the mg/d
values from Table 6.1-21 should have been mg/kg/d.
28. Figure 6.1-16, Page 6.1-135. The shaded areas
referred to in this figure are not shown.
29. Section 6.2, Environmental Protection, Page 6.2-2.
In the discussion of the applicability of subpart B of 40 CFR
191 to the FSAR, it is not clear how WIPP can become a
disposal facility without demonstration of compliance
beforehand. The FSAR will be incomplete until such demon- -.
stration has been achieved, and this should be clearly
stated.
30. Section 6.2.1.3, Non-Radiological Environmental
Surveillance Program, Page 6.2-7. This section should be
expanded to indicate the extent to which environmental
studies, if any, will be made of the ecological system to
support the assumptions and conclusions in the FSAR concern-
ing RCRA requirements.
31. Section 6.3.2, Occupational Medical Program, Page
6.3-4. Please provide additional information on this
program, such as how employees are informed about the
program, particularly the termination medical examinations.
32. Section 6.4, Industrial Hygiene, Pages 6.4-1
through 6.4-5. While this section has been expanded from two ---.
to five pages, there is nothing WIPP-specific in the section,
neither are there any references. This section is inadequate
in describing the potential industrial hygiene problems at
WIPP or the program being developed to control them. More
WIPP specific detail is needed to demonstrate that industrial
hygiene problems have been evaluated and control is assured.
Please include in this discussion an indication of how RCRA
requirements will be interfaced into the industrial hygiene
program. Also references to documents which support the
discussion should be added.
C. Appendix 6A
EEG believes this draft of Appendix 6A is seriously
flawed. There are tvo major errors of logic in the calcula-
tions. The values that are calculated are not reflected in
appropriate tables in Chapter 6. Also, there are non-
conservative changes in assumptions from the previous draft
FSAR that are not justified.
One principal problem is that the procedure used to
calculate the radionuclides present in the surface contamina-
tion is incorrect if one starts from the assumption (which
you did, and EEG agrees with) that @@the internal content of
the drum would also tend to reflect the radionuclide
distribution on the external surfaces of the respective
container." The average drum contains 658 alpha radio-
activity and 358 beta plus gamma radioactivity. Yet, the
calculation method assumes that the beta plus gamma contamin-
ation limit (which is nine times the alpha limit) is reached
first and depresses the maximum alpha contamination. The
final result is that, from an alpha/beta plus gamma ratio of
1.86 in the drum, the calculation ends up with a ratio of
0.024 on the surface! Because of a different radionuclide
distribution in boxes, the Appendix 6A methodology is only in
47
error by about 3%. However, because alpha radiation - delivers most of the internal dose, the overall dose would be
about 90% higher than calculated with the Table 6A-2 and 6A-3
values.
Furthermore, it does not appear that the values
calculated with the Appendix 6A methodology are used in
Chapter 6. For example, the 6A calculation indicates a
normal Pu-239 concentration of 6.63-16 p~i/cm3 near the drum
from resuspension, while the value in Table 6.1-5 is 6.1E-14
pci/cm3. Appendix 6A does not explicitly state how one goes
from resuspended activity from a drum or box to the average
concentration for the year. Apparently, it is assumed that
the resuspended concentration endures for one hour for each
contaminated drum or box. The number of workers exposed
during each incident is not stated. Likewise, there is no
indication of how long the resuspended concentration from
damaged containers is assumed to persist. If it is assumed A
that all 24 workers are exposed to the "6A concentration" for
1,900 hours per year and six persons were exposed to damaged
containers for 20.4 hours/year (which seems low), then the
dose would still be 0.30 + 1.00 = 1.3 person-rem/year
committed effective dose equivalent (CEDE). But Table 6.1-10
presents a value of 0.66 person-rem. To further confuse the . issue, the concentrations in Table 6.1-5 would result in a
CEDE of 0.54 rem per worker-year of exposure. It is noted
that DOE/WIPP 88-012 estimates about 10.3 person-hours of
handling for each trailer and this would be about 3.4 person-
years/year near enough to containers to receive external
radiation doses. This would result in a dose of 1.8 person-
rem/year (CEDE) from Table 6.1-5. We conclude that the
estimated doses in Table 6.1-10 cannot be reproduced from
assumptions given in either Chapter 6 or Appendix 6A and are
probably low.
Another fundamental error in methodology is the use of
the resuspension factor to calculate the amount of radio-
activity being discharged to the atmosphere from surface
contamination. Using the assumptions in Appendix 6A, one can
calculate that the total amount of radioactive contamination
on containers brought on-site during a year would be about
2.8E-05 PE-Ci. Yet, Table 6.1-12 indicates that 1.2E-03 PE-
Ci are released to the atmosphere in storage exhaust and
about O.3E-03 PE-Ci/y is released into the Waste Handling
Building. The total in Table 6.1-12 is then almost 55 times
the amount brought in! Similarly, the amount of Pu-239
reported as being released in Appendix 6A (1.9E-5 Ci/y) is
about 4.7 times that calculated in Tables 6A-2 and 6A-3. The
resuspension factor cannot be used to determine amounts lost
from a contaminated surface over a period of time because it
includes a fraction that is being continuously deposited (as
well as suspended material being transported from the
location).
The "Assumptions Usedu table (Table 6.1-4) corresponds
to Table 6.1-5 in the earlier FSAR draft. However, two key
assumptions are less conservative compared to the previous
draft:
1) the assumed number of contaminated containers
received during a year is only 52: and
2) the assumed number of damaged containers received
per year is only 192.
What is the basis for these reductions? Are there data
from waste generation and storage facilities to justify them?
CHAPTER7
Accident Analysis
A. General Comments
1. This chapter failed to adequately respond to several of
EEGvs previous comments. For example:
(a) There is still no indication that a formal Design
Basis Accident (DBA) has been performed. There are requirements
for such a DBA assessment in DOE Order 6430.1, Chapter 1. This
assessment should be performed and summarized in Chapter 7.
(b) Contamination of the underground by releases from
several accident scenarios in the CH-TRU portion of the WHB from
ventilation-air flow down the Waste Handling Shaft should be
assumed. Of even greater probability would be the transport of
contamination off-site by workers, visitors or equipment. Such
incidents have occurred in nuclear facilities on several - occasions over the years. Because of the difficulty of detecting
alpha particles this could be of particular importance at
facilities like WIPP.
(c) We had also recommended that some of the events of
moderate frequency be considered for a drum loaded with the
maximum PE-Ci level. This comment was ignored with no explana-
tion or justification for retaining the waveragel* loading. Also
see comment 2 below.
_. -- (d) Concerning Accident C2, Drum Drop from a Forklift
in the Inventory and Preparation Area, we had recommended that
; 100% credit not be taken for safety features of the facility and
equipment, and for worker training. Instead, no change was made
in the assumptions, and the exposure nallowedw is to a worker
located in a remote location. What if the forklift operator is
injured or stunned by the falling drums, or trapped and fails to
immediately leave the scene? Therefore, we still consider this
scenario to be insufficiently conservative. -
2. Although the W W has assured EEG the 1000 PE-Ci upper limit value would not be adopted until we had resolved our
differences, the WIPP Waste Acceptance Criteria has incorporated
it and it appears to be becoming a de-facto limit. The EEG has
objected to this limit since 1985, and remains opposed to such a
high value. For example, significant comments have been made by
EEG on September 27, 1985 (see comments on Chapter 7 of the SAR),
November 1, 1985, and June 22, 1988. Most of these comments are
still applicable. The basis for our objections include the
following:
(a) Such a limit permits a drastic increase in the
consequences of the scenarios presented in the FEIS. The need to
limit accident consequences from the newer inventories to those
estimated in the FEIS was the principal reason why the WW
developed the PE-~i concept in early 1983.
(b) The occurrence of the C2, C3, C4, and C6 opera-
tional accident scenarios with a 1000 PE-Ci container would
result in a committed effective dose equivalent of 400 to 700 rem
to a worker. The effective dose equivalent delivered in the
first year would be about 42 rem. Those are unacceptable doses.
The FSAR avoids presentation of this problem by assuming that
these scenarios (each assumed to occur once a year for 25 years
for a total of about 100 accidents during the lifetime of WIPP)
will always occur with a container with the average
concentrat ion.
(c) The results of Accident C10 indicate a committed
effective dose equivalent to the maximum off-site individual of
1.7 rem. Dose commitments of 3.9 rem to the lung, 29.8 rem to
endosteal surfaces, and 6.5 rem to the liver would also occur.
These doses greatly exceed a maximum dose of 0.5 rem to any
organ of an off-site individual. NRC regulations (10 CFR 60) for
a high-level waste repository require "important to safety"
structures, systems, and components to prevent or mitigate
accidents that could result in a one-time off-site dose greater
5 1
than 0.5 rem. The FSAR has concluded that there are no items -.
important to safety at the WIPP facility, consequently accident
doses off-site from the WIPP facility are allowed to exceed those
from a high-level waste repository. The effect of a serious
accident could be greatly reduced by significantly lowering the
PE-Ci limit.
(d) If adrum containing very high PE-Ci concentra-
tions were intercepted by a human intrusion borehole, there could
be more curies of TRU reaching the surface than would be
permitted by the EPA standard. It wouldbe possible to produce
about 7,000 drums of newly-generated waste at SRP and LANL during
the lifetime of WIPP with an average of 470 Ci/drum. The
probability of hitting one of these drums would be lbout 0.1.
(e) Hydrogen gas generation is likely to limit the
concentration of radionuclides in most waste forms that can be
transported in the TRUPACT to much below 1000 PE-Ci per con-
tainer.
. (f) We don't believe W E should encourage the - production of newly-generated waste that may approach
concentrations of 1000 PE-Ci per container and are unaware of any
need to do so. If the intent is not to encourage the creation of
such containers, why does W E insist on such a high limit?
In summary, we believe that the 1000 PE-Ci limit has to be
significantly lowered. Even with a somewhat lower limit it may
still be necessary to address related problems in some other way, ?" d "
such as limiting the number of high PE-Ci drums, imposing more _ " restrictive operational procedures for these drums, etc. We
would be pleased to meet with representatives of the WPO to
discuss such options.
B. Detailed Comments
1. Section 7.2.1, Source Term, Page 7.2-1. The assumption
is made that for accidents expected to happen once a year the .-
average waste package radioactivity will be used. This is not
sufficiently conservative. There should be a consideration of
the dose workers could receive from high curie packages, which
are likely to be involved in some of the approximately 100
accidents (once each year for C2, C3, C4, and C6 over 25 years)
estimated to occur. This assumption also renders the maximum
permitted PE-Ci content irrelevant for operational accidents.
The FSAR should calculate doses to workers from the C2, C3, C4,
and C6 accidents with high-curie containers, including the
maximum PE-Ci limit that is finally established.
2. Section 7.2.2.1.2, Dose Assessment, Page 7.2-4. Please
clarify whether the described use of the Rupp equation adjusts
for the acute angle release of the effluent (45 deg.), and the
forced exit-direction. If not, the adjustment should be made.
On page 7.2-5, it is difficult to see the logic in excluding
scavenging from consideration. In Chapter 6, a constant scaveng-
ing coefficient was used for estimating non-accident exposures,
and although there was some question as to the validity of using
this approximation, it was not specifically excluded. The
discussion here should address the possibility of an accident
during a precipitation event assuming that an average of about 4
such events occur per month throughout the year at WIPP (U.S.
Climate Atlas), with the greatest number of events occurring
during the summer. Also it should be made clear whether i 6
4 b resuspension of deposited radionuclides and/or saltation-creep- ya @ 1 rainsplash contamination were considered. The latter phenomenon
' B !.$ *y '\-/ is particularly important in affecting plant surface contamina-
tion in arid environments. As noted in an earlier comment under
Chapter 6 (Comment 13), the deposition velocity assumed for the
WIPP site appears to be low by a factor of two. The basis for
the selected value should be more clearly documented.
3. Section 7.2.2.2, Doses to Individuals Inside the
Facilities, Page 7.2-6. Consideration should be given to doses - to non-radiation workers inside the facilities. It is possible
that a significant exposure could occur within the fence from a
release from the exhaust stack.
4. Section 7.3.1, Incidents of Moderate Frequency
Involving CH-TRU Waste, Page 7.3-1. We concur with the con-
clusions in the CO scenario that accidents involving the unopened
TRuPACT-I1 in the radiological control area would be less than
the hypothetical accident tests and no release would be an-
ticipated.
On pages 7.3-2 and 7.3-3, two changes were made in the C2
Accident scenario. One clarified that only one drum was assumed
to be breacfied. The other updated the average PE-Ci value for a
drum. However, EEG's two main concerns (use of an average drum
instead of maximum, and assumption of maximum exposed worker)
were not addressed. It is not sufficiently conservative to 1
assume the forklift operator has left his position in about 6
seconds and thereby receives no dose. Ten seconds of inhalation
from an average drum would result in a committed effective dose
equivalent of about 38 rem, whereas a 1000 PE-Ci drum would
result in a dose of 2900 rem. This illustrates that very f i e , significant doses are possible from handling TRU waste, and this
/& 1 t fact needs to be recognized when establishing a maximum PE-Ci
limit and operating procedures for drums. The location of the
maximum exposed worker and the assumption of the average drum
also applies to the C3 scenario.
EEG comments on Amendment 9 of the SAR objected to the low
assumed fractions of the damaged drum's waste contents that were
assumed to be aerosolized and respirable (1.25E-05.h this case).
We still believe they are non-conservative by a factor of 2 to 5.
We don't want to resurrect this issue except to note that the
release fraction used should not be claimed to be a conservative -
- assumption to offset the non-conservative assumptions of an average drum and the maximum exposed individual being 20 feet
away.
5. Section 7.3.1, Accident C4, Page 7.3-5. In the second
paragraph on this page, it is assumed that the depletion of the
released activity is 20%. Please provide the basis for assuming that this is a conservative value.
6. Section 7.3.2, Limiting Incidents Involving CH Waste,
Page 7.3-7. The C8 (hoist cage drop) scenario is still listed as +--- "not credible" and is not evaluated. There has been a long
standing difference of opinion between EEG and the WPO about the
credibility of this event. In 1985, the WPO produced calcula-
tions indicating the probability was about 1.7E-08 per year ^ I-"
("Probability of a Catastrophic Hoist Accident at the Waste
Isolation Pilot Plant," THE-063). - However, a December 1987 draft report ("Quantitative Fault
Tree Analysis of the Waste Isolation Pilot Plant Waste Hoist
Hydraulic Brake Systemn), prepared as part of the Operational
Readiness Review, evaluated this same system and concluded that
V h e total probability of a catastrophic accident of the Waste
Hoist is 1.OE-03 per year (or one failure expected in 1000 years
of hoist operation)." This draft report goes on to assure the
reader that the suggested modification (providing a solenoid-
operated emergency dump value) reduced this probability to
5.2E-08, so failure is still a "not crediblen event.
We are not sanguine about the assurance .the probability is
now 5.2E-08. The 1987 analysis concluded that the 1985 analysis
was in error by a factor of 60,000. How can we be expected to
now accept the 5.2E-08 value as reliable? EEG still insists
that the C8 hoist drop accident be considered credible and the
..- dose consequences of it be evaluated in the FSAR.
It is noted that when using the assumptions for the C8 accident in the FEIS and assuming a load of seven 7-Packs, each
with the maximum thermal load permitted in TRUPACT I1 (40 watts
for two 7-Packs), one calculates a release to the environment of
1.15 PE-Ci. By extrapolation from the C10 Accident in Table 7.3-
1, this would result in a maximum off-site dose of 3.9 rem
(committed effective dose equivalent).
Concerning the C9 Accident on page 7.3-8, EEG had urged in
its October 1988 comments on the draft FSAR that the procedures
necessary to protect against a diesel fire be incorporated into
the "WIPP Standard Operating Procedure Manual," and that the FSAR
reference these procedures. This comment was not responded to in
the discussion of the C9 Accident. It is essential that these
procedures be formally adopted and rigorously followed during the
life of the facility.
7. Table 7.3-1, Page 7.3-13. The internal inconsistences - in this table that were commented on by EEG in October 1988 have
been corrected (except that the second footnote is no longer
applicable). The doses in the table can be obtained by using X/Q
factors of 2 .OE-05 s/m3 for the maximum individual, 1.7E-05 s/m3
at the site boundary, and 1.3E-05 s/m3 at Mills Ranch. The X/Q
value at the location of the maximum individual dose is only 1/3
of the 50% frequency 1-hour value from the FEIS. The X/Q values
for the other locations are 20% to 402 of the 5% frequency 1-hour
values in the FEIS. Without a detailed reevaluation of both of
these computations, it is not possible to judge which values are
the more appropriate. However, the value for the C10 accident is
high enough to be considered "important to safety" as defined in
the FSAR and the NRC regulations without using the higher X/Q
values.
8. Table 7.3-2, Page 7.3-15. The doses presented in this 1.
table are correct for the assumptions described in Chapter 7,
however they are misleading because they apply only to an
average drum and assume the forklift operator is not exposed.
Assuming a 1000 PE-Ci drum, the C2 scenario dose would be 400
rem (committed effective dose equivalent). For our assumptions
discussed in comment 4 above, the forklift operator would
receive 2900 rem.
9. Section 7.4, Accidental Releases and Exposure to
Hazardous Wastes, Page 7.4-1, 7.4-2. These accidental releases
are related to the same scenarios used for radionuclide releases,
which is a reasonable approach. The assumption that all VOC head
space gases would be released in an accident seems quite likely
because of the properties of VOC's. However, it also seems
likely that-a fraction of the hazardous components of the drums
which are not in gaseous (or volatile) form would be released.
Therefore, it is recommended that these releases be revised to
assume a fractional release of the hazardous components in a
manner similar to the fractional release of the radionuclides.
The SAR has always assumed that radionuclides would be released
from the waste matrix following an accident and there is
experimental evidence to indicate that such releases are likely.
The release fraction being used in the draft of the FSAR is
1.253-05 of the container's waste contents in an aerosolized and *
respirable form except for the C10 fire scenario which uses
2.53-03 to the drift and 5.OE-04 to the environment. Therefore
it seems appropriate to assume that all hazardous constituents in
the containers will be released from the waste matrix in the same
proportion as the radionuclides. Also, the assumed concentration
of the constituents in the drums should be that in the highest
waste form category rather than the average. These assumptions
will increase the VOC release by 20% to 701 except for the fire
scenario, where it would be increased by factors of 5 to 28. The
lead release would be increased about 2.5 times for those
scenarios other than the fire scenario, and 100 times for the
57
fire scenario. - On page 7.4-2, third paragraph, either the ground receptor
concentration or the 30 minute intake value for lead is in
error. A person would inhale about 0.6 m3 of .air in 30 minutes and if the air concentration was 5.463-12 g/m3 the intake would
be about 3-33-09 mg.
10. Table 7.4-1, Releases and Exposures from Projected
Accidents During WIPP Facility Operations, Page 7.4-3.
(a) We agree with the release values for the VOCts
(based on the assumptions) except for Freon. The Freon values
are high by a factor of about 14, however, the amount of Freon
inhaled by both the worker and off-site individual is consistent
with the correct release value (assuming that the fraction of
release inhaled is the same for all VOC's). It is of interest
that the No-Migration Variance Petition, February 1989, DOE/WIPP
89-003, Table 5-10, has a consistent value for Freon. .-
(b) Based on the assumptions presented, the lead
release and inhalation values cannot be reproduced. The value of
"potentially vaporized leada is not given. Also the fraction of
release that is inhaled is only about 1.5% of that for the VOCts
after adjusting the amount removed by filters and that plated
out. Credit should not be taken for the HEPA filters being in
operation and reducing a release, because credit is not taken in
other accident scenarios due to the fact that the filtration
system is only operative following an alarm and therefore may not
engage in a timely manner. Also a lead release could occur
without releasing enough radioactivity to trigger an alarm.
(c) The fraction of the release assumed to be inhaled
by a worker is unusually high, about 1.32 of the total release.
The fraction for radionuclide inhalation in this chapter is less
than -012 of the release. Also Table 5-10 in the No-Migration
Variance Petition referred to above shows a worker's intake to
be only 5.33-032 of the release. Therefore, the 1.3% inhalation -
value is obviously in error and should be corrected.
(d) The effective X/Q values at the location of the
maximum off-site individual for the VOC releases are 1200 times
greater than the X/Q values used for the radionuclides. Please
explain the basis for this difference.
CEAPTER8
Long Term Waste Isolation Assessment
A. General Comments
1. In our October 14, 1988 review of the Draft FSAR, EEG
strongly objected to the deletion of 159 pages of detailed
discussion, summary, and tabulations estimating the consequences
from long-term waste isolation and pointed out that this violated
the 1 9 8 1 DOE/State Consultation and Cooperation Agreement which
specified content of the SAR. This Draft FSAR responded to our
objection by reinstating 18 pages of summary statements,
conclusions and tabulations heavily referencing Amendment 9 of
the SAR.
This re-insertion of consequence analyses in Chapter 8 by
reference to an earlier SAR could be claimed to have resolved the - issue of non-compliance with the C&C Agreement. However, it is a . superficial outdated effort. None of the tabulations and figures
are more recent than March, 1983. There have been drastic
changes in the inventory since that time and the method of <
calculating radiation dose has been changed. EEG considers this b '
l long-term Waste Isolation Assessment to be inadequate.
2. Failure to Meet 1 9 8 1 CLC Aareement
The text begins with the statement "The purpose of this
Chapter is to -cuss the long term isolation assessments that
will a ~ ~ l y to the WIPP facilityn (emphasis added). While the
text does provide a minimal discussion, that is not the purpose
mutually agreed upon by DOE and New Mexico in the July 1, 1 9 8 1
Stipulated Agreement, Appendix B, Working Agreement that specified the contents of the Safety ~nalysis Report, Chapter 8,
Long Term Waste Isolation Assessment. It was agreed that the -
document would analyze the long term impact on public health and
safety following decontamination and site control termination
and would include consequence analyses. This chapter does not
discuss the 1981 commitment by W E to perform consequence analyses or even reference the Working Agreement.
3 . Failure to Provide Co m ~ w a b l e SAR to WE HL W SAR
DOE has agreed to complete their SAR for the HLW repository in Nevada before they begin construction of the repository. That
SAR will include an evaluation of the performance of the proposed
geologic repository for the period after permanent closure and
give the rates and quantities of releases of radionuclides to the
accessible environment as a function of time and a similar
evaluation which assumes the occurrence of unanticipated
processes and events. Why can the Departmant agree to provide
this detailed information in their SAR for HLW in Nevada and not
provide it in the FSAR for TRU waste in New Mexico? Note that
A this issue is independent of whether the facility is a repository or is a research and development facility. Both are analyses of
the safety of a proposal to place unwanted radioactive materials
in a mine.
4. It is stated that until the decision is made regarding
the use of the WIPP facility as a permanent repository, com-
pliance with Subpart B of 40 CFR 191 is not required. The
following reasons are advanced for not demonstrating compliance
at this time:
-"." > Possibility of Revision of Subpart B by EPA. I :' . .
," ;< : ,: : "wt ?; ~ !.' ,.. c,
{ :, ; . ~ t ":: j ! . Further experiments and analyses are needed to complete
a performance assessment.
Since it is not expected that there will be any major
changes in Subpart B, and an agreement with New Mexico to adhere
to the vacated standards is in effect, anticipation of a revision, is not a justified reason for non-compliance. The FSAR should be
more specific as to the analyses involving the collection of
data. What is the specific data that must be collected to refine
assumptions? What are these assumptions? If undefined assump-
tions have been formulated, then they should be stated in this
report, as well as any supporting analyses. If experiments are
to be performed, then they should be described and schedules
presented.
5. The issue of demonstrating compliance with Subpart B of
the EPA Standards for the disposal of transuranic waste and
performing long term waste isolation assessment of consequence
analyses is separate and administratively unrelated as the
following chronolagical sequence indicates.
7/81 W E agrees to conduct consequence analyses in the
SAR (Ref. W.A.) -,
11/84 W E agrees to meet any future EPA disposal
standards (Ref. 1st Mod. )
9/85 EPA promulgates standards for disposal of TRU ,. -, : waste.
Since there is nothing in the C & C Agreement and subsequent
modifications to relieve the Department of its obligation to
conduct these SAR Analyses, on what basis does the Department
contend that the obligation to demonstrate compliance with
Subpart B of 40 CFR 191 relieve W E of the 1981 SAR obligation?
B. Detailed Comments
1. Section 8.1, Sumamry of Initial Consequence Analyses - 62
Performed for WIPP, Page 8.1-2. This section references
Amendment 9 of the Safety Analysis Report. Since the FSAR will
supersede all previous amendments to the SAR, it does not seem
reasonable to adopt or take credit for passages in earlier
versions by reference. As previously recommended, the long-term
consequence analyses should be included in the FSAR, Chapter 8.
2. Page 8.1-2, "These standards now exist in 40 CFR
191...'1 - Subpart B of the standards does not now exist. "The WIPP facility must demonstrate compliance to these new
standard^...^^ - They were promulgated in September 1985 and are not new.
3. Section 8.1-12. While the issues of compliance with
the EPA standards and performing consequence analyses in the SAR
are mutually unrelated events, the discussion of EPA standards
contains a number of misleading and incorrect statements.
The text states that compliance with Subpart B of 40 CFR 191
is not required until the decision is made to use the WIPP
facility as a permanent repository. Two reasons are provided for
not demonstrating compliance at this time:
A) Possibility of Revision of Subpart B by EPA, and B) Further experiments and analyses are needed during the Pilot
Plant Phase to complete the performance assessment.
Neither are correct.
With respect to A), a formal agreement exists between New
Mexico and DOE to evaluate the expected performance of the
proposed repository with the vacated standards. Hence, the
possibility that the standards may change is not germane.
Additionally, anticipation of a revision of the standard is not a
justified reason for non-compliance, particularly when all
parties agree that most of the standard will be salvaged. With
63
respect to B), to date there are no experiments nor analyses that - have been identified that are needed for performance assessment.
4. Section 8.1, Tables 8.1-1-6. It is noted that all
doses in these Tables are from Amendment 6 (March 1983) or
earlier versions of the SAR. The inventories used have been
changed significantly. Also, old dose conversion factors and the
pre ICRP-26 & 30 method of dose calculation is still used.
5. References for Section 8.1. Other than a 1986
revision of the SARI the remaining references are 1978 or
earlier.
6. Section 8.2.1, Performance Assessment, Page 8.2-3. As
stated in our earlier comments, DOE-WIPP 86-013 requires that
sensitivity and uncertainty analysis be carried out. This
section again fails to recognize the need to include uncertainty
analysis in the performance assessment methodology. Also, the
final scenario report has not been published as of April 1, 1989.
7. Section 8.2.1.1, Scenario Development and Screening,
Page 8.2-4. As previously recommended, the discussion on human
intrusion modeling should include consideration of the pos-
sibility of Castile brine reservoirs under the repository. (See
EEG reports EEG-11 and EEG-15).
8. Section 8.2-3, 8.2.1.1, "A final scenario report will
be published in 1988..." - It is now May 1989, the report has not yet been published, and the future tense should not be used to
describe a 1988 publication date.
9. Section 8.2-2, "Activity to address each of the
assurance requirements is scheduled to begin in FY88..." - FY89 is now over half over and the sentence should be rewritten as,
"Activity m.. ." or in FY89. -
lo. The following paragraphs "Summarize the progress to
date8@ - What progress has occurred since September 19851 Nothing
but a schedule has been published.
Section 8.2-7, 8.2.1.5. Our October 14, 1988 comments on
the internal Peer Review Panel have been ignored and are
reprinted again.
11. The text states that W E is the only implementing
agency responsible to determine compliance with the standards and
an internal Peer Review Panel will provide assurance to state
officials can be assured that WE's conclusions are credible.
That philosophy virtually guarantees a loss of credibility with
the New Mexico EEG if the intent is merely to ask us to review
the results. The authors appear unfamiliar with the 1978
contract between W E and the State of New Mexico.
12. Section 8.2-10. The schedule shows completion of
Subpart B compliance in October 1992. No indication is provided
of the amount of time between completion by DOE and review by EEG
and others. Contrary to the text, the Compliance Strategy (Plan)
does not provide such a schedule.
13. What does %ajorn and "supportingn mean in the diagram?
For example, will scenario development be completed in April
19901 That is not consistent with the plan to publish scenario
development before October 1988.
Conduct of Operations
A. General Comments
1. This chapter contained several significant improve-
ments, and evidences considerable response to previous EEG
comments and recommendations.
**Although DOE is responsible for all aspects of the WIPP
facility, it delegates those functions to various contractors.**
The chapter is silent on the responsibility to protect the
workers and the general population except for identifying the
Safety, Security, and Environmental Protection Department as
being responsible for **health and safety related programs which
satisfy the requirements of the DOE and other...agencies ." The philosophy and tone of the responsibilities and authorities of - the various officials do not convey a strong commitment to health
and safety matters.
B. Detailed Comments
1. Section 9.1.1, Owner Organization. The text states
that the functions, responsibilities, and authorities of W E and
its contractors are discussed in Section 11.1.1. They are not.
2. Section 9.1.2.2.1, Page 9.1-2, General Manager. The
General Manager has overall responsibility for the operation,
maintenance, 'and modification of the WIPP facility. Is the
General Manager ultimately responsible for the health and safety
of WIPP personnel or has this authority been delegated to a lower
level?
3. Section 9.1.2.2.7, Page 9.1-4, Safety, Security, and - 66
Environmental Protection Department. This paragraph assigns the responsibility of health and safety to the "department." The department should have llfunctions,w and responsibilities should
be assigned to an individual, such as the Department Manager.
ihis comment also applies to other sections in the chapter where
the "departmentn is assigned responsibilities.
4. Figure 9.1-1, Page 9.1-8, Management & Operating
Contractor Organization Diagram. The management diagram does not
reflect a communication line between the General Manager and the . .
. . Radiation Safety Manager. The WIPP "Radiation Safety Manual," WP ' , . , , , ?.
: ' : j 12-5, assigns the responsibility for interpreting the radiation , , . ~ . . . .
.. : : .: safety program to the Radiation Protection Manager, yet the FSAR, ?:: , , , :
- ' Section 6.1.5.2, assigns responsibility for the radiation safety
program to the Safety, Security, and Environmental Protection
Manager.
The radiological safety program responsibilities should be
clearly defined and reflected in the formal organization
structure. Please clarify the reporting and communication lines.
5. Section 9.1.3.2, Page 9.1-6, Staff Managers. Although
there is reference to staff manager's qualifications, there are
no requirements specified. The words -typically haven should be
replaced with "as a minimum requirement shall have." The
importance of qualifications should be reviewed with respect to
guidance found in ANSI/ANS-3.1-1987, "American National Standard
for Selection, Qualification, and Training of Personnel for
Nuclear Power Plants." Although this document is not a general
M)E requirement, the overall guidance should be followed at WIPP.
Chapter 9 should state commitments to high level management
qualifications, and specifically to appropriate technical
experience of the Radiation Protection Manager. As per ANSI/ANS-
3.1-1987, the collective qualifications of management and
technical managers shall be reviewed and supplemented, as
necessary, with personnel with applicable qualifications. - 6. Section 9.2, Acceptance Testing, Page 9.2-1. This
section refers to the "WIPP Procedure Manual.I1 Presumably this
is a reference to "Standard Operating Proceaures,I8 WIPP-WE-103.
If so, this title should be correctly presented. EEG does not
have a document entitled "WIPP Procedure Man~al.~~
7. Sections 9.2.2, 9.3.5, and 9.4.1, Acceptance Tests,
Administration and Records, and Plant Procedures, Pages 9.2-3,
9.3-5, and 9.4-1. These sections refer to Sections 11.1.11,
11.1.12, and 11.1.17. There are no such sections. They should
refer to Sections 11.11, 11.12, and 11.17.
8. Section 9.4.4, Operational Occurrences, Page 9.4-2.
The text discusses compliance with U.S. DOT regulations in the
transportation of TRU wastes, but fails to discuss compliance
with U.S. NRC regulations. A.
9. Section 9.4.4, Operational Occurrences, Page 9.4-3.
This section refers to DOE Order 5484.2 which has been super-
ceded by W E Order 5000.3.
10. Section 9.4.4.1, Page 9.4-3. The assumption that a
contaminated drum, box, or canister would not contaminate the
interior of the Internal Containment Vessel on the TRUPACT may
not be valid and can result in reduced worker safety.
CBAPTER 10 Operational Safety Requirements
A. General comments
1. There have been substantial improvements in this
chapter, and it has been responsive to many of EEG's comments and
recommendations.
2. The introduction states that RH-TRU waste handling is
not covered and that, "This document will be expanded to include
those OSRs (Operational Safety Requirements) prior to receipt of
RH-TRU." EEG agrees that this supplement will be necessary.
B. Detailed Comments
1. Section 10.1, Introduction, Page 10.1-1. We disagree
that, "It is inconceivable that non-radioactive hazardous
materials would be released from containers without the simul-
taneous release of radioactive material^.^' Our reasons include:
(a) the VOC's are much more volatile than the trans-
uranics ;
(b) in some containers there may be heavy concentra-
tions of hazardous chemicals and low amounts of radioactivity;
and
(c) the amount of radioactivity released may not be
enough to trigger an alarm.
This assumption should not be made before sufficient
operational experience is obtained to verify it.
2. Section 10.1.4, Definitions and Acronyms, Page 10.1-4.
The following acronyms should be added: AC (Page 10.6-4,5), OSR
(Page 10.6-3) .
-. 3. Section 10.3.1.1, Continuous Air Monitors, Page 10.3-2.
This section implies that only two CAMS are mandatory, and
therefore the CAMS used for Effluent Monitors, as discussed
under Section 10.2.1.2, are not required. This is incorrect.
Also, the title of this section should be amended to "Continuous
Air Monitors for Waste Handling B~ilding.~
4. Section 10.3.1.2, Effluent Monitors, Page 10.3-4. The
second paragraph under LC0 should be amended to include activity
alarm limits for the Station B CAM. Since Station B represents
filtered exhaust from the storage horizon, it should have the
same alarm limits as Station C from the Waste Handling Building.
Such an alarm system would provide an alert to defective
filtration in the event of a release followed by an alarm at
Station A (also see discussion on page 10.3-11). Furthermore,
there is a reference to the LC0 for Station B under "Appli-
/'- ~ability" on page 10.3-5. -
(=' ' ? I! On page 10.3-5, it is recommended that this discussion be > q $ L * , damended to indicate that portable equipment would only be
<- """' acceptable for use at Station C if it were connected to the
isokinetic probe. The use of batch sampling for monitoring of
this effluent point should be used only as a last resort, and
for a very short time period.
5. Section 10.3.2.1, Waste Handling Building Differential
Pressures, Page 10.3-6. Normal differential pressure ranges for
the four WHB areas were given in the first Draft FSAR, but
deleted here. Why? Is the system in the WHB able to meet these
previously mentioned differential pressures? On page 10.4-2 this
draft still takes credit for listing the pressure differentials.
6. Section 10.3.2.3, Underground Exhaust Air Filtration
System, Page 10.3-11. The last paragraph on this page discusser-
the HEPA filtration system provided for the underground exhaust.
It states that periodic verification of the efficacy of the
filters is required to maintain confidence in their ability.
This discussion should refer to Section 10.4 for information on
how verification will be provided. section 10.4 indicates that
the filters will be verified by local examination at each shift.
A local visual examination may not be sufficient to determine
that a filter system is ineffective. A,more definitive descrip-
tion of "local examinationn is needed here. At the present time,
there is no alarm if the exhausted air exceeds prescribed
radiation limits; therefore, the filters could be defective
throughout an entire shift, or longer, if the "local examinationn
is not adequate. See comment 3 above.
The CAU at Station A would be sampling a significant
dilution of the radioactive particulates if there should be a
release. Consideration should be given to initiating filtration
of the underground effluent based upon alarms from CAMS located
inside the RMAs of the underground. Dilution of the contaminated
air would be less and the air being monitored would be compara-
tively free of interfering salt dust.
J /-"--* w , 7. Section 10.4.1.2, Effluent Monitors, Page 10.4-2. This
$,section also uses the ambiguous phrase "local examination." This , , 4
i phrase should be more definitive. For example, it could refer to
a specific WIPP procedure. In Section 10.4.2.3, the requirements
indicate that the filter banks will be tested only annually.
Therefore, it is essential that the 8vlocal examination" can
actually determine effectiveness of the effluent monitors.
8. Section 10.4.2.3, Underground Exhaust Air Filtration
System, Page 10.4-4. This section requires only annual verifica-
tion of the effectiveness of the HEPA filter banks. Because of
their importance, it would seem desirable to increase the
frequency of such verification. Please provide the basis for
such infrequent verification. - 9. Section 10.5.4, Ventilation Systems, Page 10.5-3. The
first paragraph on this page states that, "All effluent air
streams from areas that contain radioactive materials are
filtered monitored for activity." It should be made clear
that the normal operating mode is to exhaust unfiltered monitored
air from the exhaust shaft. All air effluents at WIPP are not
filtered.
10. Sections 10.6.1, 10.6.2 and 10.6.3, Training, Design
and Procurement, and Document Control, Pages 10.6-2, 10.6-3 and
10.6-4. The references listed on these pages should include the
"Radiation Safety Manual," WP 12-5.
11. Section 10.6.4, Audit Program, Page 10.6-4 to 10.6-6.
The references listed on this page should include the "Radiation
Safety Manual," WP 12-5, since there are several important limits - which are set forth only in this manual.
12. Section 10.6.8.2, Area Radiation Monitors, Page 10.6-
10. This section allows the ARMs to be reset to higher levels
for an indefinite period if a higher radiation source is in the
area. This seems to defeat the purpose of the ARMS and could
allow indiscriminate violation of the established limit of 10
mr/hr. The section should be revised to more definitively
establish criteria for resetting to a higher level, and limiting
the time at which it may remain at the higher level. Also
resetting to a higher level should be permitted only if author-
ized by a health physicist.
CBlLPTER 11 Quality Assurance
A. Detailed Comments
1. Table 11.1, Applicable Quality Assurance Standards,
Page 11.1-4. Part B of this table and associated discussions
should be revised to include reference to ANSI/ASME NQA-2,
Current Editions, nQuality Assurance Requirements for Nuclear
Power Plants." NQA-2 is applicable to the operations phase of
work at nuclear facilities and is to be used in conjunction with
applicable portions of ANSI/ASME NQA-1.
2. Section 11.2.2, General Responsibilities, Page 11.2-5.
Documentation should be added to describe the implementation of
the Quality Code Classification work described in Guidelines for
Requisitions to Determine Quality Code Classification for
Purchase Requisitions and Purchase Requisitions Change Notice
Attachment 2 of Westinghouse Procedure 15-009, Revision 2.
Documentation should also be added to describe the Quality
Surveillance work required by the Westinghouse Procedure 13-011,
Revision 0, Quality Assurance Surveillance.
CBAVPW 12 -~ Decontamination and Decommissioning of the WIPP Facility
A. General Comments
1. Chapter 12 assumes that the facility will meet the EPA Standards for disposal of TRU waste during the five-year
demonstration period and that the removal of 65,000 drums will
not be required. A safety analysis should reflect conservative
assumptions on matters affecting the health and safety of workers
and the general public. Hence, this chapter should contain plans
and safety analyses of potential radiation doses to workers and
the public from operations and transportation if the wastes need
to be retrieved, returned to the generating sites, sent to the
high-level waste repository, left indefinitely on the surface at
WIPP, sent to a new site, or left in place.
- B. Detailed Comments
1. Section 12.1, General, Page 12-2. In the first
paragraph of page 12-1, the reference to "DOE Order 5280.2An
should be 5820.2A.
2. Section 12.2, Decontamination and Decommissioning, Page
12-3. This section lists the sequence of future planned events
for decontamination and decommissioning, but does not provide
meaningful information either in this section, or the chapter,
to permit a safety evaluation of the processes. Additional
detail is needed for a safety analysis.
3. Section 12.5, Post Closure Physical and Environmental
Surveillance, Page 12-5. As previously indicated, additional
detail is needed. For example, further information should be -
- included on how mining will be controlled. Furthermore, the
plans for surface environmental surveillance do not appear to
address the intent of the NRC in 10 CFR 60 which requires
subsurface early warning detection. The National Academy of
Science report also recommended subsurface surveillance.
APPENDIX:
OCrOBER 1988 CO-S
- ENVIRONMENTAL EVALUATION GROUP - U1KIYw.rruyn/- IM-mO1B1a
7007 WYOMING BOULEVARD. N.€ SUITE F.2
October 16, 1988
M r . Jack B. Tillman Project Manager VIPP Project Office U. S . Department of Energy P. 0. Box 3090 car1sb.d. m 88221-3090
Dear Mr. Tillman:
Our reviev, vhich contains 77 pages of detailed comments and recomended changes, of the d r a f t DOE F i n d Safety Analysis Report (FSAR) is attached. Several of the SAR chapters ref lected a conscientious response by the YIPP Project Office (VW) t o EEG's December 1986 and January 1987 co-nts on Amendment 9. The Final SAR, however, does not contain an adequate analysis of the safety of the VIPP pro jec t as required by the Ju ly 1. 1981 Working Agreement for Consultation and Cooperation. as well a s DOE Order 5681.1B. A considerable amount of work needs to be documented and risks quantified before the SAR can be considered acceptable.
W e have repeatedly requested tha t the SAP. include a fo r th r igh t statement of the purpose of the VIPP project . vhich is designed t o serve as a repository fo r permanent i so l a t ion of defense transuranic vaste from the environment. Also, the purpose of the SAR has been dovngraded from .. . . the wst comprehensive document . . . on public health and s a f e t y . . .' ( W E m C 6 C Agreement) t o *. . . t o support the construction and operation of the VIPP.' (Draft =All, page 1.1-1)
The 'draft f i n a l g document contains serious def ic iencies , misleading i n f o r u t i o n , ooissioru of su f f i c i en t de ta i l , f a i lu re s t o respond t o our previous c o m n t s . and minor e r rors . The w r e s e r i o w deficiencies involve primarily Chapter U, 6 . 8. 10, and 12. Chapter U f a i l s t o address the c r i t e r i a used a s a bas i s fo r the sumary d conclusions; Chapter 6 does not respond t o our previous coDpents on Amendment 9 a s you agreed t o do i n your
nr . Jack 8. Tillman October 14, 1988 Page 2
l e t t e r of 3ovember 19. 1987; Chapter 8 f a i l s t o provide a long-term waste i s o l a t i o n assessmenc a s requi red by the 1981 DOEINew Mexico Consul ta t ion and Cooperation Agreement and t h e 1988 DOE Order DOEIAL 5b81.18; Chapter 10 needs t o provide considerably more d e t a i l t o permit u s t o eva lua te t h e o p e r a t i o n a l s a fe ty ; and Chapter I2 should conta in d e t a i l e d plans f o r decommissioning and deconcaminacion, i f r e c r i e v a l of nuc lea r waste becomes necessary .
Due t o the substancfve na tu re o f our o b j e c t i o n s :o t h e F ina l SAR, EEG r e q u e s t s recons idera t ion of t h e p o s i t i o n taken i n your September 15, 1988 l e t t e r t ha t you w i l l not respond t o our couments p r i o r t o pub l i sh ing t h e F i n a l S a f e t y Analysis Report. Your e a r l y response t o t h i s r eques t is needed t o i n s u r e t h a t our concerns a r e addressed.
- Robert H. Nei l1 Di rec to r
RHN:EU: l s b
Enclosure
cc : M r . James Bickel. A s s i s t a n t Manager M O . WE Albuquerque. New Mexico
D r . Laurence Lattnm. P r e s i d e n t Nev Mexico Tech Socorro. New W i c o
ENVIRONMENTAL EVALUATION GROUP
m A N ~ ~ I . S C . U . W ~ ~
7007 WYOMING BOULEVARD. N E. SUITE F.7 - ~ -
ALBUQUERQUE. NEW MU(IC0 87104 1505) 8281003
Comments and Recommendations
on
WE Draft Final Safety Analysis Report
Waste Isolation Pilot Plant
Prepared by
Environmental Evaluation Group
New ~ C X ~ C O
October 1988
The preliminary Safety Analysis Report was first published by the Department
of Energy's WIPP Project Office WPO) in 1980. The New Mexico Environmental
Evaluation Group (EEG) reviewed this five volume document shortly after
publication and transmitted detailed comments and recommended changes to the
WPO. Following such review and comments, and on some occasions, meetings.
between representatives of respective agencies, the report has been revised
nine times. These amendments have been based upon both W E initiated changes
in design, criteria. and p l a ~ e d operations. as well as EEG recommendations.
The present W E draft Final Safety Analysis Report reflects a favorable
- response to many of EEG's previously recommended changes, but there remain
some serious deficiencies, some misleading information. omissions of
sufficient detail, failures to respond to previous comments, and minor errors.
The more serious deficiencies are summarized below, with the more detailed
observations and recommended changes itemized under each Chapter heading.
1. The SAR does not provide an adequate safety analysis of WIPP including the
quantification of risk and documentation of calculations and analyses
envisioned in the 1981 Consultation and Cooperation (C & C) Agreement
\ between DOE and the State of New Mexico. Even the purpose of the SAR has
been downgraded from .the w s t comprehensive document . . . on public - -
health and safety . . .. agreed to by DOE and the State in the 1981 C & C
Agreement to the present 'to support the construction and operation of
the UIPP. *
- 2 . The purpose of the UIPP Project is to provide a facility to permanently
isolate the defense TRU waste from the environment. It should be clearly
stated in the first paragraph of Chapter 1 of the FSAR. Statements such
as. "The UIPP facility is designed to receive, inspect containers for
damage and contamination, emplace. and store unclassified defense-
generated transuranic wastes in a retrievable fashion . . ." (Chapter 1.
page 1.1-1) and "The UIPP is designed as a full-scale facility to
demonstrate the technical and operational principles for the permanent
isolation of defense-generated transuranic waste in salt* (Chapter 3 .
page 3.1-1) are misleading.
3 . The draft FSAR violates the 1981 M)E/State Consultation and Cooperation
Agreement (Appendix B. Article 111). and the requirements of the 1988 WE -.
Order AL 548l.lB. both of which require the FSAR to have a long-term
waste isolation assessment, identifying potential communication modes.
modeling methods, and consequence analyses. Earlier drafts of the SAR .
,F &::,& . . ..,. . , . . did attempt to evaluate long-term vaste isolation (159 pages in the ~..'>. * , ;. " "
; , $ > 5 .,: . , , , , , . , previous edition), but such analyses have been replaced in the Final SAR . ".. . . .,
v. ; ., % ' .. , * . . ,# ,/ with a nine page general discussion of how perfonunce assessment will be ... ..̂
conducted. Comequence analyses Q U S ~ be included.
4. Although the Department has informally indicated that there is no intent
to conduct experiments at UIPP with high level waste, there are numerous
references in the SAR to high level waste experiments. They should be
deleted since the SAR contains neither technical justification nor
radiological risk evaluation of bringing high level waste.
5 . The SAR takes credit in Chapter 8 for a Peer Review Panel providing
assurance on suitability of WIPP as a repository. Since the Department
has never involved EEG with any of the Peer Review Panels, nor provided
us with agenda, minutes or recommendations, we believe that the
committees do not provide credibility as stated in the SAR, but in
actuality detract from it. In order to take credit, EEG must be
involved.
6. According to the August 4, 1987 Second Modification to the C b C
Agreement. DOE agreed to document compliance with Subpart A of &O CFR 191
in the Final SAR prior to the receipt of waste. Such compliance has not
been adequately shovn nor specifically identified.
7 . Chapter LA is important because it is intended to provide a summary of
.ore detailed information contained in other chapters for final
--. conclusions on the safety of the WIPP design, criteria and plans for --,.,
I . . .a_
' . , , , . " s i 1
S; .. operations, and for long-term stability. Unfortunately, Chapter U fails . . .: ;il. ! . , .! '.i 1
, , '1 :< i / x.; , , , . ',. to present the basic criteria used in cooing to the conclusion that WIPP
. . , I , ... .. . 2, can be operated safely.
8. While the SAR invokes frequent comparisons with the NRC safety regulations
for nuclear reactors. it would be far more appropriate to make
comparisons with the NRC health and safety standards for high-level waste
disposal (10 CER 60).
9. It is noted that all references t- backfilling the waste storage rooms
have been deleted from the FSAR. We know that the Departoent has decided
to emplace the experimental CH-TRU waste without backfill to avoid
crushing the drums during the retrieval period. The FSAR should clearly
state this decision and the reason for it and should state that when the
waste is emplaced for disposal, a properly selected tailored backfill
will be used to fill the space between the druns, above the drums and
between the walls and the d w s . The FSAR should also state that only
the mount of waste expressly needed for conducting experiments to help
in Performance Assessment (to show compliance with the EPA Standards
40 CFR 191 Subpart B) will be emplaced in a temporary mode without
backfill.
10. While WE made a conscientious response to EEG's December 16, 1986
comments on Amendment 9 of the SAR, an important exception vas Chapter 6,
Environmental Safety and Health Protection, in which it appeared that the , . . ._\
: ' >, "'. $ ' : j .; A , 1. .i. . ; , authors were neither aware of our previous comments nor the commitments
% ." '-~ '2& J +*'" contained in the November 19. 1987 response by the UIPP Project Manager.
Chapter 6 needs to be extensively revised by addressing our coments.
1 The Operational Safety Reviev (Chapter 10) lacks sufficient detail to
pernit us to evaluate the operational safety of VIPP. EEG's specific
comments describes the areas in vhich extensive expansion and revision
are nee&d.
12. Since it is not apparent that a Design Basis Accident (DM) assessment
has been perforwd, it is strongly recomncnded that such an assessment
addressing the requirements of W E Order 6630.1 and the guidelines to "A
~uide to Radiological Accident Consideration for Siting and Design of W E
Non-Reactor Nuclear Facilities." U-l0296-~C, be performed and summarized
in Chapter 7.
13. Chapter 12 assumes that the facility will meet the EPA Standards for
disposal of TRU vasce during the five year demonscraeion period and chac
the removal of 125.000 drums will not be required. A safety analysis
should reflect conservative assumptions on matters affecting the health
and safety of workers and the general public. Hence, this chapter should
contain plans and safety analyses of potential radiation doses to workers
and the public from operations and transportation if the wastes need to
be retrieved, returned to the generating sites, sent to the high level
waste repository, left indefinitely on the surface at UIPP, sent to a new
site, or left in place.
CHAPTER 1
Introduction and General Description
1. Section 1.1, Page 1.1-1. The purpose of 'the WIPP project to provide a
facility to permanently isolate defense TRU wastes from the emrironment is
not stated anywhere in this document. We understand that the Department's
data needed for the analysis to show that the project can accomplish that
purpose is not yet available, however, the purpose should be clearly
stated in the document. Statements such as on page 1.1-1, "The VIPP
* . i... . , . facility is designed to receive, inspect . . . " aremisleading and raise
'8: $\ '': ,' unnecessary questions as to the intent of constructing the repository.
2 . The text states. "The final SAR has been prepared by W E to support the - construction and operation of the WIPP . . . .' This stated purpose
appears to be considerably less comprehensive than that agreed to by WE
and New Uexico in Article 111 of the Working Agreement For Consultation
and Cooperation (1981) vhich states, T h e Safety Analysis Report (SAR), as
amended from time to time, constitutes the most comprehensive document
concerning UIPP both in general and specifically as related to public
health and safety u well as other matters.' Use the definition in the
C 6 C Agreement since W E and New Mexico devoted four pages in that
Agreement to delineate the expected scope of the SAR. It would also be
helpful to state in this section that the Final Safety Analysis Report
(FSAR) is required by the 1981 State/WE Consultation and Cooperation
Agreement, and the 1988 W E Order 5081.111, and its purpose is to provide
analyses of the risks associated with the location. design, and operation - of the facility, a s-ry of applicable standards and criteria, and to
demonstrate through such documentation that it meets applicable standards
and requirements for public health and safety.
3. The last line of the first paragraph of this section should be amended to
indicate that the facility will not become a permanent disposal facility
through "successful demonstration of the acceptability." but through
successful demonstration of compliance with the EPA Standards 40 CFR 191,
followed by a decision by W E to make it a permanent disposal facility. . ." 3 i: 4 ; : , ; i.4 '
> . ,, 9;. 6,) , , ., i, \ < '& ;!
. . .. .. j I.:,! $ r 4. ~ l l references to the high level waste experiments at UIPP should be
%. . t' deleted from the SAR since WE has not developed any plans for such
experiments nor intends to do so.
5. Section 1.1.1. Page 1.1-2. A reference to the presence of "three mines
are located between five and ten miles of the site" should be expanded
with a description of lease holdings at the VIPP site and the distance of
nearest mining activity from the edge of the site. A =re recent and
reliable estimate of resources at UIPP than the information contained in
the FEIS is a report by the New Mexico Energy and Uinerals Department.
entitled *Natural Resources at the Waste Isolation Pilot Plant," January
1984.
6. Section 1.1.2. Page 1.1-2. Please enunciate the 'technical and
operational principles" to be demonstrated. Describe the 'studies and
experiments . . . to extend the understanding of the behavior of high- level waste in salt' or cite a reference where these are described.
Describe the .operating and scientific &tag that is expectad from
temporary emplacement of waste or provide reference to a document that
describes it. Finallyi provide the basis(es) on which 'WE will make a
decision regarding whether to dispose permanently of transuranic waste at
VIPP." The role of the State of Nev Mexico in this decision should be
indicated since the text implies that it is a unilateral decision by WE.
Section 1.1.3. Page 1.1-3. The text tmproperly states that the CH-TRU
6 storage capacity is 6.3 x 10 cu ft based on the assumed drum-box split.
6 The capacity of 6.3 x 10 cu ft is specified in the 1980 VIPP FEIS.
While the text states. 'A specific VIPP storage capacity has not been
established for RH-TRU waste," the C 6 C Agreement (Modification 1,
11/30/84) specifies that the limit of RH-TRU waste is 5.1 x lo6 curies.
The current concept of disposal of decommissioned, contaminated surface
facilities of VIPP should be expanded to include plana for the disposition
of these contaminated facilities.
Section 1.1.3. Page 1.1-0. "The design a c c o d t e s the time required
to reach the waste and retrieve it if such a decision is ma&.' Please
provide specific time periods for the emplacement and retrieval of each
waste form.
Section 1.1.3. Page 1.1-4. The definition of when UIPP becomes a
permanent waste facility on page 1.1-4 (after decommissioning) differs
from the definition on page 1.1-1 (upon successful dmomtration of the - acceptability). Actually. VIPP becomes a permanent facility when the
waste is emplaced with no in ten t of recovery (40 CFR 191.02 1). Ihe SAR
should describe plans fo r review and approval by EPA. EEC. and the State
of New Kexico, before WIPP becomes a permanent f a c i l i t y .
12. Section 1 .1 .4 , Page 1.1-4. The schedule should be amended t o r e f l e c t
recent changes. Also. please provide a basis for the plamed schedule.
13. Section 1.2.3.3, Page 1.2-6. Please ident i fy the location of the waste
experimental area and provide a br ief description of these experiments.
14. Section 1.3. 2nd Paragraph, Page 1.3-1. This paragraph discusses the
. possible need for allowing exception t o the WIPP Waste Acceptance
Cr i te r ia . The Environmental Evaluation Croup (EEC) has par t ic ipa ted
exteruively i n reviews of the or iginal UAC and the various revis ions. We
recognize the importance of the c r i t e r i a to the safety of WIPP operations
and to public health and safe transportation of WIPP wastes. Therefore.
it is our firm conviction t h a t the State and EEC should be informed i n
advance of proposed revis ions o r exceptions t o the WIPP WAC, and that we
should be provided the opportunity t o make a determination of the safe ty
impact independent of the VW. This policy would be consis tent with I,
% Amendment 1 of the WE/State Consultation and Cooperation Agreement,
which s t a t e s .. . . the WE w i l l , pr ior t o granting such exceptions f o r
such waste and prior t o the shipment of such waste: (1) perform analyses
t o ascer ta in the impact of such on the public health and safety. (2)
consult with the State of New Mexico, including providing the S t a t e with
a copy of the analyses fo r review and comment, and (3) provide t o the
Sta t e a period of for ty- f ive (45) days t o review and coment on such
analyses prior to granting any such exceptions. In no instance v i l l such -.
an exception t o the WAC be granted i f it would cause a s ign i f i can t
increase in the impacts on public health and safety discussed i n the VIPP
FEIS." The EEG i s not i n s i s t i ng tha t such a policy be s t a t ed i n the
FSAR, bur it should be made a par t of the Standard Operating Procedures
(SOP) for amending o r alloving exceptions to the VIPP WAC. The f a i lu re
t o include such a policy in the SOP would represent a f a i l u r e t o adhere
t o the s p i r i t of the comitment re f lec ted i n the VIPP WAC and the C & C
Agreement.
..;. ' . \ .,* k., . Also, it would be desirable t o include i n t h i s sect ion a reference t o the ,-,' ' 3 :. ;$ need t o w e t the requirements of 40 CFR 191. . ., ., +.-'
15. Section 1.3.1. P a ~ e 1.3-2. I t is noted tha t a11 references t o
backf i l l ing the vaste storage rooms have been deleted from the FSAR. We
knov tha t the Department has decided t o emplace the experimental CH-TRU
vas t e without backfi l l t o avoid crushing the drum during the r e t r i e v a l
period. The FSAR should c l ea r ly s t a t e t h i s decision and the reason fo r
it and should s t a t e t h a t when the waste is enplaced fo r disposal. a
properly selected ta i lored backf i l l w i l l be used t o f i l l the space
betwean the drums. above the drums and between the malls and the drums.
The FSAB should also s t a t e t h a t only the amount of waste expressly needed
f o r conducting experiments t o help i n Performance Assessment ( t o shov
compliance with the EPA Standards 40 CFR 191 Subpart 8) m i l l be emplaced
i n a temporary mode without backf i l l .
.- 16. Section 1.3.G. Page 1.3-3. This section on defense high level waste
experiments should be deleted since there are no plans to conduct such
experiments.
17. Section 1.5.1, P a w 1.5-1. The second paragraph indicates that site
characterization was completed in December 1978. As indicated .n the
title of a Sandia Report (SAND 88-0157) published in b y 1988 (received by
*~.. -- EEC on July 1, 1988). 'Summy of Site-Characterization Studies conducted , C ' -*
"-* ~, ,.)? . ., -: 4 ::: , +.,
from 1983 through 1987 at the VIPP Site." site characterization is , % ,. i : .. - , '
, , ' ongoing. Also, Mr. Troy E. Wade 11, Acting Assistant Secretary for ,
Defense Programs, in his testimony before the Subcommittee on Procurement
and nilitary Nuclear Systems of the Committee on Armed Services of the
U. S. House of Representatives on September 8, 1988, said. 'Formal WIPP
site characterization activities will end in December 1988.' Only the
first phase of site characterization was completed in 1978. The first
sentence on page 1.5-2 is 8 description of the site characterization
studies performed since 1983, but the attempt appears to have been made to
not use the phrase 'site characterization* and thus the description is
meaningless. We suggest using the appropriate language from the s-ry
of SAND 88-1057 to describe this effort.
18. Section 1.5.2. Page 1.5-2. 1.5-3. It should be made clear that the
experimental programs described in this section do not require emplacement
of any radioactive vaste. In order to avoid confusion. it would be
appropriate to clearly separate the R and D that requires waste from
experinents that are performed to better &sign the repository.
19. Section 1.5.2.1. Page 1 .5 -3 . The discussion of the t e s t s t ha t have been - performed for several years should include a t l e a s t the preliminary
r e su l t s of those t e s t s . What have the plugging and sealing s tudies shown
,. t o date? Uhen is the decision on plugs and seals design expected to be ,/'
i i S,\ . , ' . made? What assumptions on plugs and s e a l s performance w i l l be used for ; yL, i; , ,,
I ..s . . ' !:j ' 1
t :* , . . . ~ . , showing compliance with the &PA Standards? Please indicate when room- ' ,.
sca le sea ls w i l l be evaluated.
20. Section 1.5.2.3, Page 1.5-5. This sec t ion does not describe experiments
with waste and, therefore , the t i t l e is nisleading. Please change the
t i t l e t o "p TRU Waste Container S t a b i l i t y . "
21. Section 1.5.2.4. Page 1.5-5. Please change the t i t l e of t h i s sect ion to
"RH TRU Waste Container S t a b i l i t y . " -
22. Section 1.5.3, Page 1.5-8. This s ec t ion re fers t o "the design
modification a l te rna t ives . ' These a l te rna t ives should be l i s t e d and the
one tha t was selected should be j u s t i f i e d . Also, describe which of the
"follov-on studies' a r e being performed and why. and include a summary of
the resu l t s t o date.
23. Section 1.5.3.1, Pages 1 .5-8 , 1.5-9, 1.5-10. Please provide references
fo r the s tudies described in t h i s sec t ion and a t l e a s t a br ie f synopsis of
the resul ts .
28. Section 1.5.0, Pane 1.5-12. A s previously indicated, please de le te - references t o high-level waste experiments.
-.
2 5 . Table 1.5-1. Pages 1.5-16. 1.5-17. New dates are needed for publishing
the "Interim Sorbing Tracer Test Report", "Hydrogeochemical Facies in the
Rustler Formation", "Groundwater Modeling Study of the Rustler Water-
Bearing Zones*, and the "Facies Variability and/or Evaporite Dissolution
Within the Rustler Formation" reports.
. ..r: CHAPTER LA ( &< 'j .. i-4%. : < , : ,*%.a " . Summary Safety Analysis 3 h<, \,, e ,::, ! " : 2 . 7 . ,,
9< $, , . k,.,
1. This Chapter attempts to summarize the entire FSAR in order to present the
conclusion that it is a safe facility. It does in fact sumarize the
radiation doses to the vorker and the public as a result of normal and
abnormal operations and accidents. It then concludes that there is no
significant impact on che public health. It fails, however, to summarize
the criteria that vould provide the basis for such a conclusion. These
also should be summarized, either as a table or narrative. For exapple,
the FSAR must provide reasonable assurance that the facility will meet
Subpart A of 40 CFR 191. i.e.. the combined annual dose equivalent to any
member of the public in the general environment resulting from (1)
discharges of radioactive material and direct radiation from such
management and storage shall not exceed 25 millirems to the whole body. 75
millirems to the thyroid. and 25 aillirems to any other critical organ.
2. Section U.1. Page 1A-1. It is recognized that Chapter 1A is a S-ry
Safety Analysis, hovever, the attempt to summarize in the first paragraph
leads to some confusion and ambiguity. For example, the definition of
Design Class I in the first paragraph is a serious distortion of the more
detailed definition in Sections 3.1.7.1.1. To refer in the first
paragraph to .a leak' froa the UIPP facility. and then discuss effluents
and controlling radiation doses off site in the second paragraph, will
almost certainly lead to confusion for some. Please also see our comnents
on the definition of Class I items in Section 3.1.7.1.1. -
3. Section l .A.1.2.1.2. Paae U . 1 - 6 . The f i r s t section describes a potent ia l
problem which would be resolved by means of an inexpensive t e s t . It is
recommended that the types of c lay minerals present i n the Dewey Lake
Redbeds be determined to resolve the question referred to .
The f in i t e element model mentioned i n the th i rd paragraph of t h i s sect ion
predicted the closure r a t e s t h a t were 3 t o 4 times slower than t h e actual
observations. This sec t ion should include the discrepancy between the
predicted and actual creep r a t e s and what changes i n the design and
operations plans have been made t o accommodate tne f a s t e r r a t e s of
closure.
4 . Section 1.A.l.2.2, Page 1A.1-7. The f i r s t sentence of t h i s s ec t ion is
- incorrect . There i s plenty of evidence of i n f i l t r a t i o n below the
Uescalero Caliche. The second sentence should be more exp l i c i t . The
+/ . ' vater-bearing zones a r e a t depths of approximately 608 t o 632 f e e t . 71& to .fl,>!
. ,pi i: Z : i * ) ., 740 feet . and 850 to 860 f e e t . I n the t h i rd sentence, because grouting of
i t , 5 . , . { 3
\ .: * . \,Ys ..,
t he shafts had to be performed th ree times before leakage vas stopped, it : ., ."... ,. , .,.'
should be indicated tha t frequent checks and maintenance v i l l be provided
t o ensure that the sha f t s reaain dry. The fourth sentence. "No
grouadvater forces i n s a l t a r e experienced because there a r e no connected
saturated pore spaces" is incorrect . (See Bredeholft. 1988; and Beauheim.
1987. page 0 . l ines 1 t o 4 . ) Final ly . the l a s t word of t h i s s ec t ion
should be "important" and not "s ignif icant . '
5 . Section 1.A.1.3. Page 1A.1-7. Poten t ia l damage from an a i r c r a f t crash
should not be indicated a s 'very ?mall." While the probabi l i ty of a
scheduled a i r l i n e r crash may be very small, a number of small a i r c ra f t and - mili tary a i r c r a f t are seen f ly ing very low over the s i t e . In f a c t , a
small private plane crash occurred a t the s i t e i n 1.982. Efforts should be
made to safeguard the s i t e from such incidents i n the future. ~ l s o .
please note our comment 5 on Chapter 2 .
6 . Section U.2.3.1. Page U . 2 - 5 . The l a s t paragraph of t h i s section
s t a t e s tha t the gamma source s t rength was used t o calculate shielding for
RH-TRU ra ther than neutron dose r a t e s . Please indicate a reference for
information which provided a bas i s fo r t h i s conclusion.
7 . Section U.2.3.2. Page U . 2 - 6 . The l a s t paragraph of t h i s section
. . implies tha t the doses t o the public fo r normal operations are developed - assuming use of HEPA f i l t e r s for the exhaust from underground. I t should
be mentioned tha t t h i s is a va l id assumption only fo r accidents.
8 . Section lA.2.b.l. Pages 1A.2-8 and lA.2-11. The uni t s used in Table
U .2 -1 a r e confusing. From Table 6.1-8, it is apparent that doses due t o
"direct radiation' (ex terna l rad ia t ion is preferred terminology) a re
average individual doses and should be labeled rem/year. The doses due t o
inhalation a re cal led population doses i n Table 6.1-10 ( t o t a l dose
received by a11 exposed persons), which would be 50 year dose commitnent
i n person-rem.
9 . Section lA.5, P a ~ e U .5 -1 . This s ec t ion s t a t e s . .The f a c i l i t y operation
w i l l include i n s i t u experiments addressing technical issues for defense 6
waste programs and s torage of defense r e l a t e d contact-handled (CH) md
remote-handled (RH) transuranic (TRU) vaste." At least a brief
description of experiments and technical issues to be resolved should be
provided to support receipt of vaste at WIPP for the R and D purposes.
References should be provided of documents in vhich a detailed description
of the experiments may be found.
...
The statement, "This section provides conclusions vith regard to the
adequacy of the WIPP site." is incorrect. Either remove the statement or
provide the conclusions vith appropriate rationale.
10. Section U.5.1. Pages U.5-I. U.5-2. Please provide appropriate
references for the "conceptual design" and the "preliminary design."
Page U.5-2, first paragraph, mentions keeping a minimum clearance for
five years. It should be changed to at least ten years since the
retrieval vould take at least as long as emplacement. The same paragraph
discusses the models to predict the rate of closures vithout mentioning
that these predicted rates vere three to four tinvs lover than the rates
observed after excavation.
Befora reaching the conclusions stated in the last paragraph of this
section, it is necessary to discuss the changes in the plans for
retrievable emplacement of vaste vhich became necessary due to rates of
closure three to four times higher than predicted. For example. it is
planned to enplace the vaste without backfill during the period vhen
retrievability nee& to be maintained to minimize the possibility of the
drums being crushed. Also, the rooma will be cut an extra foot higher and
trimmed before emplacement of waste. The original design criteria were. - therefore, not suitable. This section should discuss vhy these changes
became necessary, what vere the options available, vhy vas an option
chosen, and why the changes are not likely to affect the operations in the
facility adversely.
Reference 1 cited at the end of this section is not listed at the end of
the Chapter. There are no references listed.
11. Section U . 5 . 2 , Page U . 5 - 3 to l A . 5 - 5 . The last paragraph on page U . 5 -
3 should be expanded to include the plans for disposal of excavated salt.
Before concluding (last paragraph of this section on page 1 A . 5 - 5 ) that,
"Under no circumstances will the public health and safety be subjected to - significant risks,' this section should provide references to detailed
information on the accident scenarios which were analyzed. and potential
releases from malfunctioning or inadequate design of shafts and
underground facilities. (Also, please note comwnt one on this Chapter.)
CHAPTER 2
Site Characteristics
In response to many EEG comments on previous versions of SAR. this chapter has
been extensively revised and is much improved. A fev questions that remain
are listed belov.
Section 2.1.1.2, Pages 2.1-4 through 2.1-5. This section refers to W E
Order 5180.U vhich has been superseded by W E Order 5480.18. Also, it is
noted that the dose limits on page 2.1-5 exceed the limits of the EPA
Standard 40 CFR 191.03(b). Under 40 CFR 191.04(a). W E may establish
alternative standards for WIPP provided EPA has approved an application
for such alternative. If W E has made such application, please provide
EEG vith a copy of the application and EPA approval notice. W E Order
5480.1B references W E Order 5480.11 for radiation protection standards.
however, this order indicates that the "Requirements for Exposure of
Individuals and Population Groups in Uncontrolled Areas" is to be added at
a later date. Therefore, if no alternative standard has been approved by
&PA, the FSAR should adhere to the exposure standard prescribed in 60 CFR
lgl.O3(b).
Section 2.1.2.1. Page 2.1-6. Please update paragraph NO of this
section since the land exchange of State-ovned sections has been
completed.
Section 2.1.2.1.2, Pane 2.1-7. Settlement of potash Leases inside the
WIPP site should be completed before a permanent decision is madm to store
wute at VLPP.
-19-
4. Section 2.1.2.2. Paaes 2.1-8 and 2.1-9. The last paragraph of this
section refers to "low level of radioactive releases . . . : Please
include references to the analyses of postulated accidents which support
this statement.
5. Section 2.2.3.1. EEG personnel have witnessed very low flying military
aircraft directiy over the UIPP site, almost on a daily basis during
certain periods in summer, 1988. Either such flights should be banned or
their potential impact on UIPP should be considered in Section 2.2.9, and
evaluated in other Chapters of the FSAR. One crash has occurred already
at the UIPP site.
6. Section 2.1.3.0. Pane 2.1-10. Tbe reference to "population centers" as
defined in 10 CFR 100 is no longer appropriate. Those regulations apply
,+. , . . to siting of nuclear reactors, not nuclear waste repositories. A more .,: ;C ,,.,,
appropriate reference would be 10 CFR 60 or 00 CFR 191. With respect to qt2.: . . ,i,'
i ir ... !. y'* , '
exposure to members of the public off-site, it would be preferable to . t... ,.. . . .. , refer to the limits of 40 CFR 191.03(b).
7. Figure 2.1-7. The location of the potash leases of the International
Minerals and Chemical Corporation are not shovn on this map.
8. Section 2.5.3.5.2. Page 2.5-23 throu& 2.5-27. This section is very
well vritten, but needs to be updated.
9. Section 2.5.4. Pages 2.5-29 through 2.5-34. This sect ion needs a
thorough updating in viev of the recent publ icat ions by Bredeholft, Novak.
UcTigue. Beauhein, e t c .
10. Section 2.5.5.1, Page 2.5-37. The l a s t paragraph of t h i s sec t ion on
page 2.5-37 refers the reader to Chapter 8 , but t he nev Chapter 8 of FSAR
is devoid of any technical evaluations.
11. Section 2.5.6, Paae 2.5-46. The f i r s t sentence on page 2.5-46 is
incorrect . In a kars t region. such as the WIPP a rea , lack of a near-
surface regional vater table does not preclude a s ign i f i can t quant i ty of
recharge. -The fac t is tha t i n s p i t e of several suggestions by EEC over
the past f ive years, the WIPP Project Office has chosen not t o perform
systematic studies t o determine the r a t e of recharge a t and i n the
v i c i n i t y of the WIPP s i t e . It is more cor rec t t o say t h a t the information
is not available ra ther than assuming it t o be 'negligible."
12. . Section 2.5.7.4, P a ~ e 2.5-58. In view of the discussion i n t h i s and the
preceding subsections, it is recomended tha t Table 2.5-21 not be included
i n the FSAR. This table and the l a s t paragraph on page 2-5-50 should be
replaced by a discussion of the program t o determine more r e l i a b l e values
of Kd being performed currently.
CHAPTER 3
Principal Design Cr i t e r i a
1 . Section 3.1, Pare 3.1-1. This section re fers t o experiments and
s tudies to understand the behavior of waste i n s a l t . Please add the
; ,.t.. " , reference to reports o r chapters i n the FSAR where these experiments are 8, ~: ..
5 ".,
'. i ' described.
2. Section 3.1.1, Paae 3.1-2. Experimental (high-level) waste a s one of
the types of waste t o be emplaced should be deleted.
3. Section 3.1.1.1.1. Page 3.1-3. This sect ion should c l ea r ly indicate
tha t the boxes described i n Table 3.1-2 a re not compatible with TRUPACT-
X I . Therefore, there should be an indication of how these boxes w i l l be -̂
transported, i . e . , whether they w i l l be repackaged in to a TRUPACT-I1
compatible box, or whether another type B container, c e r t i f i e d by NRC
w i l l be used. Also, Table 3.1-2, page 2. should be revised t o indicate
the type of "standard waste container' t o be used.
4. Section 3.1.1.1.2. Page 3.1-3. Radionuclide composition of waste for
VIPP has been calculated and presented i n Table 3.1-4, based on a March
1987 report (Ref. 4 ) . To date . the VIPP of f ice has refused t o provide us
with th i s report. In order f o r EU: t o conduct its evaluation, the report
must be provided.
5 . Section 3.1.1.1.2. Page 3.1-4. This sec t ion s t a t e s t h a t the average -.--.
Pu-239 equivalent is 7 g f o r a drum and 110 g fo r the most co-n box.
Table 3.1-3 shows 16 g of Pu-239 in drums and 79 g in .standard waste
containers." It is recommended that this section, or Table 3.1-3, be
revised to reflect consistency and the basis for the values presented.
Section 3.1.1.1.6. Page 3.1-5. This section also should indicate that
the thermal power will be limited by the PE-Ci limits and the Watt limits
of TRUPACT-11. Please also add a reference to the data which reflect the
fraction of heat
EEG information.
the total volume
UIPP.
source plutonium to be shipped to UZPP. According to
the SRP heat source wastes alone comprise about 102 of
and 602 of the total alpha curies that will be cooing to
Section 3.1.1.2, Page 3.1-5. This section is misleading in that it
implies a very small fraction of the UIPP waste is RH-TRU. State that
RH-TRU waste comprises 362 of the total radioactivity.
Section 3.1.1.2.2, Page 3.1-6. The first paragraph of this page
contains incorrect reference des4.pations. On line 2, the reference to
the current data base probably should be 6 rather than 7. Also, the
reference designation in the last line should 8 rather than 9.
Section 3.1.1.2.3. Page 3.1-6. This section also designates 9 as
reference for surface dose rate limits, and it should designate 8 (the
first modification of the C & C Agreement).
Section 3.1.2.2. P a m 3.1-7. In this section, it vould be helpful to
state hov many TRUPACTS could be stored outside and how many drums inside
(in both surge storage and shielded CH-TRU area) and the basis for the - limitation.
11. Section 3.1.3.1, Page 3.1-9. This section states that the RH through-
put would be two canisters per shift, which indicates a throughput of 500
canisters per year, which is greater than 250 canisters per year. Uhich
is the maximum?
12. Section 3.1.3.3, P a ~ e 3.1-11. In the fourth line from the bottom. the
reference designation should be 9 rather than 10. Also, the designation
. . ..' .,,. . i ;: -,
10 CFR (clfii) is incomplete and, therefore, aobiguous. It should . . . .
f / ,, r : . I ! :. * >; <:, , , ; : , < ' . .,! ,%,
designate vhich subparagraph of (c) is applicable, i.e., (c)(2)(ii) or .. . . , 4 : ~ .; ,,' . . , ' ,,; . i .
(c)(6)(ii) or both.
-.
13. Section 3.1.3.3. Pane 3.1-12. The reference designation in the first
paragraph is in error. Please see previous comment.
1-5. Section 3.1.3.3, Page 3.1-13. The second paragraph on this page
describes movable shielding in the floor hatch which mates vith the
bottol of the facility cask to allow access to the cask loading room
while the cask is being loaded. Is there r control or alarm provided to
preclude access when the shield is not in place? If so, this should be
described here.
15. Section 3.1.6.2. Page 3.1-16. This section indicates radioactive waste
generated on-site vill be disposed at UIPP. Compliance vith the EPA -.
disposal standards n u t be completed befor* any vaster can be disposed at
VIPP. Please provide the design for solidification of liquid wastes at
UIPP. EEG has not received any information on this.
16. Section 3.1.7.1.1. Pages 3.1-17 to 3.1-18. The first paragraph defines
the Design Class I items and relates these items to a "basic component"
as defined in 10 CFR 21.3 (a) (2) as follows:
.,."--'~ "'Basic Component,' when applied to other facilities (than nuclear power {.' g::,
reactors) and when applied to other activities licensed pursuant to Parts [ 4 g ? ?* ,,. . , . ?
30, 40. 50. 60, 61. 70, 71, or 72 of this chapter, means a component. I . . :
system or part thereof . . . in which a defect . . . could create a
substantial safety hazard . . . . " "A 'substantial safety hazard' means
a loss of safety function to the extent that there is a major reduction
in the degree of protection provided to public health and safety for any
facility licensed . . . pursuant to Parts 30, 40, 50, 60, 61, 70, 71, or 72 of this chapter.'
The only one of the listed parts of Title,lO which would have provisions
somewhat comparable to the WIPP facility would be Part 60, 'Disposal of
High Lvel Radioactive Wastes in Geologic Repositories.' In that Part.
Section 60.1 defines 'important to safety" ax follows:
"'Important to safety,' with respect to structures, systems, and
components means those engineered structures, systems, and components
essential to the prevention or mitigation of an accident that could
result in a radiation dose to the whole body, or any organ, of 0.5 rem or
greater at or beyond the nearest boundary of the unrestricted area at any
time until the completion or permanent closure:
- 25-
- Therefore, the def in i t ion of Class I i n t h i s paragraph of the d r a f t FSAR
is nor consistent with 'a basic component" as defined i n 10 CFR 21. AS
EEG has previously maintained, the Class 1 def in i t ion should be amended
so t h a t it is consis tent , and so that it w i l l not appear that the design
c r i t e r i a for UIPP provides l e s s protection than t h a t which w i l l be
imposed on WE for high-level vaste repository. Perhaps the reason tha t
the VPO has res i s ted t h i s change is because there would then e x i s t some
Class I items a t UIPP, j u s t a s there a r e l ike ly t o be some Class I items
a t a high-level geological vas te repository. If not , then there should
be no reason for not making the two def ini t ions ident ica l .
17. Section 3.1.7.1.1. Paae 3.1-18. The third paragraph on chis page
s t a t e s that "each WIPP item was evaluated against the design .., ...
.,* ': c lass i f ica t ion c r i t e r i a . . I t is presumed tha t such evaluations are
i'dii. $5.
1, "'1' ' '. documented, however, no reference is c i t e d t o t h i s documentation. Please , ,'
provide th i s reference i n order that EEG and others may obtain copies.
18. Section 3.1.7.3. Page 3.1-21. This section lists several design
fac tors vhich provide a b a s i s f o r the design c l a s s . Among these is
includod the phrase .importance t o safety; vhich is not defined. I f
t h i s phrase also has the same meaning as i n 10 CFR 60.1. it fur ther
substantiates the need t o modify the Class I de f in i t i on as discussed i n
comment 11 above. I n any case, t h i s phrase should be defined and i f it
i s not the same as i n the NRC regulations for high-level waste
repositories, thc EEC vould lib t o knov the r a t i o n a h f o r the
difference.
- 19. Section 3.1.7.5.1. Paae 3.1-22. The cited reference 13 in this section
should be changed to 11.
20. Section 3.1.7.5.2, Page 3.1-22. The cited references in this section
are also in error.
21. Table 3.1-3. Page 3.1-27. This table has two errors, as follovs:
. a. Th-282 should be Th-232, . c j : ; ; , ,, >..., . . . , .
.I 5,:
b. The activity for Pu-238 should be l.OE+Ol .__I,",'
22. Table 3.1-3, Page 3.1-28. This table also has two errors, as follovs:
a. The Ci/container of Pu-239 should be 4.9 E+OO.
b. The total activity should be 83.
23. Table 3.1-5, Page 3.1-30. If the average canister contains 1000
curies. then the first modification to the C 6 C Agreement (vhich limits
total curies in BH-TBU to 5.1 million) will limit the number of canisters
to 5100 and to 162.000 cubic feet of vaste. rather than 250,000 as stated
in Section 3.1.1.2.3. It is also noted that the VIPP Final Environmental
Impact Statement (FEIS) shows 510 Ci/canister. 170 Ci of Cs-137. 160 Ci
BA-1371, and 145 Ci of Pu-241.
24. Section 3.3.1.1. Page 3.3-1. The last sentence on this page is - misleading. A significant fraction of the UIPP waste may be mobile since
the Waste Acceptance Criteria (WAC) allow one percent of the particulates
to be less than 10 microns in particle size.
25. Section 3.3.3. Page 3.3-17. Paragraph (1) of this section indicates
some 'minor exceptions" to the general rule that 'all buildings and their
support structures are protected by fixed, fire suppression systems . . .
." Please add to this paragraph some examples of such exceptions. Of
particular interest is whether any fixed permanent structures lack such
fire suppression systems.
26. Section 3.3.6. Pane 3.3-19. This section discusses radiological
protection design considerations. In the second paragraph, it states
that the plant is designed so that under normal operating conditions.
"the radiation exposure to the general public is negligible as compared
to the natural background radiation.. It then references section
6.1.6.4. This reference section does not clarify the waning of
"negligible,* but instead merely discusses the effluent and environmental
monitoring program. For the FSAR, it is necessary that the criteria
forming the basis for the 'negligible' conclusion be clearly stated. and
that this section should refarence tho relationship bem~een potential
doses to the public and the design criteria. These criteria and
relationship are clarified for occupational exposure, but not for the
general public.
CHAPTER 4
Plant Design
1. This chapter contains many improvements and a considerable amount of
additional important information. With only a few exceptions, the
-.-. -% information also was well presented. The few exceptions concerned some
of the figures which were not always legible in their newly condensed
form.
2. Section 4.2.1.5, Pane 4.2-15. The last paragraph of this section
states the liquid waste produced at UIPP will be solidified prior to
disposal ;t UIP~. Additional details should be added here on the plans.
designs, metho& of processing, and radiation protection procedures. since
it could be a factor in evaluating radiation safety.
3. Section 4.2.2.3, Page 4.2-19.. The first paragraph erroneously refers to
Figure 4.2-7 for the air intake shaft headframe. It should reference
Figure 4.2-6.
4. Section 4.2.5. Page 4.2-20. This section refers to Figure 4.2-10 for a
display of the pumphouse and contents. Figure 4.2-10 is a pictorial of
the east and south elevations of the Support Building. There is no
figure in the 4.2 series which displays the pumphouse.
5. Section 4.2.2.6, Pam 4.2-20. The first paragraph refers to Figures
4.2-12 through 4.2-15. The correct references are Figures 4.2-7 through
4.2-10. There are no Figures 4.2-12 through 4.2-15.
6. Figure 4.2-4. Page 4.2-30. This figure is an unacceptable
representation of the configuration and operation of the EFB for the
following reasons:
a. The effluent monitoring station locations are improperly identified.
The reduction in size and extension of the duct at Station B is not
represented. Station A is nov co be down the exhaust shaft, not
after the bend as shown.
b. The norplal operating position of the dampers of the EFB is improperly
,/ ,., :- ., identified, e.g., the EFB isolation dampers should be normally t . .
i *.% -< ~,
"closed" not "open.' to allow normal discharge to occur by the new , ,
. % , - main fans (which, incidentally, are not identified in the figure). -_
7. Section 4.3.2.1, Page 4.3-10. This section describes the subsurface
facilities and refers to Figure 4.3-5. This figure does not indicate the
location of the area for plug-in battery charging. The plug-in area
should be indicated since it is a location of potential fires or
explosion should ventilation be irudequate.
8. Section 4.3.2.1.2. Page 4.3-9. This section states that vapor from
diesel fuel constitutes the principal risk of an underground explosion.
Other areas warranting consideration are those for vehicle battery
charging as discussed in Section 4.3.2.1, and the waste storage area when
filled with waste. The latter deserves consideration unless the emplaced - waste will be. covered with backfill following emplacement. Audits of the
U I P P Waste Acceptance Cr i te r ia do not ver i fy absence of pyrophoric wastes
or those materials which tend t o produce explosive gas mixtures.
9 . Section 4.3.2.2, Page 4.3-14. Please provide a reference t o the
procedure to be used for r e t r i eva l of contaminated and uncontaminated
waste packages. Such procedures a re e s sen t i a l t o an adequate radiat ion
safe ty analysis.
10. Section 4.3.2.3, page 4.3-14. In the l a s t paragraph on t h i s page.
there i s a discussion of the dose r a t e from emplaced RH-TRU. I t is
/ . s t a t ed tha t the maximum loaded canis te r vould be 100 remhr. This is not . , . - 8. i 2; $ i . + i J . . , : . . . correct . Pursuant t o Modification 1 of the DOE/State Consultation and
\.^ . ' '
" Cooperation Agreement, up t o 5% of the RH-TRU canisters nay have a
surface dose of 1000 remfir. Also, please provide the reference o r more
d e t a i l i n the l a s t paragraph on t h i s page on the calculation tha t
concludes the highest radiat ion dose r a t e around the shield plug w i l l be
only 3 mrenfir.
11. Section 4.4.1.2.1.2, Page 4.4-13. The first l ine on th i s page
indicates t ha t ' i f excessive airborne contamination is detected a t the
a i r intakes, the outside a i r supply is d i rec ted through HEPA f i l t e r s . "
I t is not c lear in Figure 4.4-7 ( the Lcgend page is not legible) i f a
separate CAH is provided for the intake. o r w i l l the a i r be monitored
elsevhere? Please c l a r i fy .
12. Section 4.4.1.3. Page 4.4-16. t h i r d paragraph. This paragraph
indicates t h ~ t 'a small quantity of a i r is drawn down the vaste
shaft . . . . " What is meant by "a small quantity'? Has an analysis been - performed of the potential radiological health consequences of an
airborne release incident in the UHB which results in Contaminants being
dram dovn the waste shaft and exposing workers below?
13. Section 4.4.3.1.2.3. Pane 6.4-34. Please provide a reference to .. ~. ..
, . '̂ . :; .. ., figures which display the location and other details of the "separate
: r ' i, .." , :,.. 3. i . , , . . . i . , ; exhaust systemw provided for venting hydrogen f r w the battery-charging !,, ,: ?j '.P 4
ti ;; ,..,,, y.$ , . , . . i' process. ' ., J'
14. Section 4.9.3.1.6. P a ~ e 4.4-37. Please include a reference to the
report of the study done to evaluate public fire fightins capability and
level of mutual aid response in case of a fire emergency at VIPP.
15. Section 4.4.3.2, Pane 4.4-37. This section refers to Table 4.4-10
which provides a considerable amount of nev and helpful information on
the fire suppression equipment throughout the plat.
16. Section 4.4.4.1, P a ~ e 4.4-47. The first paragraph of this page refers
to Figure 4.2-10 for a schematic of the pumphouse. As indicated in
coment G hove, thir is an incorrect reference. There is no figure
shoving the pumphouse.
17. Section 4.4.6. Paae 4.4-51. The second paragraph of thir section
refers to Section 5.4 for details on process operations, however, Section
5.4 fails to provide details on the solidification process for liquid
waste. Aa previously stated, info--tion on this solidification process
is necessary for an adequate safety evaluation. The Failure and Affects
Analysis (Table 4.4-13) does not consider this subject. Furthermore.
such processing should be clearly defined prior to the receipt of WIPP
wastes.
18. Section 4.4.10.4, P a ~ e 4.4-7L. This section describes the breathing
air supplies available for fire fighting and normal WIPP operations.
More detail should be added on the equipment and procedures used for
producing the Class D breathing air for SCBA cylinders. Information also
is needed on the inventory of SCBA tanks. No breathing air compressor
systems are indicated.
19. Section 4.5.2.5. P a ~ e 4.5-6. This section describes the canister
shuttle car used to 'transfer waste canisters from the port in the floor
of the hot cell to the floor of the cask loading room. Please add a
scale drawing of the shuttle car.
20. Section 4.6.5, Page 4.6-2. More detail is needed on the criteria to be
used to determine when reinforced canopies will ba used to protect
operators of -bile mining equipment.
CHAPTER 5
Process Description
1. Page 5.1-1. If DOE decides to ship waste by rail to UIPP. EEC requests
six months prior notification in order to address education of the public
and rail workers to potential doses and problems.
2. Section 5.1. Page 5.1-1. Following the sentence, .The shipping
containers are DOT Type B containers certified by the NRC: please add the
sentence "These containers are designed for shipments exceeding 20 curies
of plutonic and are. therefore, doubly contained.'
3. Table 5.1-3, Accidents 15 and 16. The stated consequences are not
consistent vith the information provided in Chapter 7, and the - -. ..
*I..
..consequences as stated for Accident 18. Hov can an accident be bounded by *.% ,
b :. E ::,, ,' \ . I .. , , another accident which is not considered credible. and therefore the 7 .",
consequences vere not analyzed?
4. Section 5.2.1.2. Page 5.2-3. The second paragraph on this page should
contain a reference to the Standard Operating Procedure to be usedfor
this and subsequent operations. These procedures are necessary for an
adequate safety analysis
5. Section 5.5.2, Page 5.5-2. The first paragraph of this section refers
to W E Order 5480.'~. Chapter V. but references DOE Order 5480.18, Chapter
XI. Please clarify. Since DOE 5480.1B superseded 5480.U. references to -.
5480.U should be deleted.
CHAPTER 6
Environmental Safety and Health Protection
1. In general, Chapter 6 appeared to have been prepared hastily. There vere
few improvements, and there was no evidence of response to EEc's
December. 1986 comments on Amendment 9 of the SAR. The Bibliography for
Chapter 6 vas poorly done, containing several duplications and
inaccuracies.
2. Section 6.1.1.1, Page 6.1-1. Credit is taken for periodic AURA
reviews of design documents by nuclear and health physics specialists
from the architect-engineer organization. EEC has not received any of
their reports, analyses or recommendations. Please provide this
information.
3. Section 6.1.1.3. P a ~ e 6.1-3. This section refers to the YIPP Radiation
Safety Manual (UP 12-5) as the basis for the UIPP AIARA programs. EEC ....
,, f * 6 , : agrees that this manual vill impact substantially on the radiation safety [ & !;! fi , ,
i, r? .~ aspects of YIPP. and we recently transmitted comments and recommendations
t,$ .< for the improvement of the policies, procedures and criteria contained in
that manual. Ye are pleased that most of our recommendations were
adopted in the July 1988 revision of the manual.
The SAR states that the UIPP AURA program is described in Section 2.0 of
the YIPP Radiation Safety Manual (UP 12-5). Section 2.0 of that document
contains only generalized responsibilities of management, persome1 and
radiation workers. Hence, the SAR does not adequately describe the WIPP
AIARA Program as claiaad and should do so.
-35-
I. Section 6.1.2.1.1. Paae 6.1-6. Please add a table in support of this
section to show the estimated nunber of 55 gallon drums of waste and the
number of each type of boxed waste to be shipped to WIPP.
5. Section 6.1.2.1.2, P a ~ e 6.1-5. This section refers to Table 6.1-2 for
the design basis gamma source strengths for RH-TRU vaste, however, this
table provides only average source strengths. The description should
include the maximum source strengths of 1000 remfhr.
6. Section 6.1.2.1, Pages 6.1-4, 5. The CH-TRU end RH-TRU gamma source
strengths presented in Tables 6.1-1 and 6.1-2 are not derived from the
current- radionuclide content of RH-TRU waste presented in Table 3.1-5.
The Mev/s-canister total in Table 6.1-2 should be less than one-quarter
of the value given, and the peak energy group area should be in 0.h-0.8
P n
! \ $3 not 0.8-1.3 Mev and 1.3-1.7 Mev. Vhile the values used probably ' : %, 8; f$
$ , ?%*. '., ; *>, .v , gf; . 3 .give consemative results for shielding and extern~l dose considerations.
:, .-,it ';.; ,?- ,
it would be desirable to directly determine the g m spectrum on
representative drums. An assumption that exterrul radiation does not
contain significant quantities of lov energy radiation can lead to the
selection of instruments or TLD badges that fail to detect the radiation
present.
The comment was made that the alloved neutron dose rate is much smaller
than the allowed g- dose rate. A table similar to Table 6.1-1 should
be provided for neutron energy group and surface flux. -4
7. Section6.~.2.2,Pal~es6.1-6throuah6.1-9. Theprocedureusedto
estimate resuspended radionuclide concentrations, quantities of
radionuclides inhaled, and the resulting radiation doses from handling
undamaged, surface contaminated containers is difficult to follow. We
also question some of the assumptions and are unable to arrive at similar
conclusions. Also. the reference to the August 19. 1981 version of W E
5480.1B may not be appropriate in view of the 1988 draft version of
5480.11. if it has been approved.
The surface contamination concentrations given in Table 6.1-4 are only
#-* '*, about 5 2 1 of the WAC limits for both drums and boxes. This will. of
! course, impact subsequent calculations. In our comments on this table in
Amendment 9, we pointed out that the values were incorrect and were told
- the table would be completely revised in the FSAR. The numbers are
unchanged from Amendment 9.
The expression on Page 6.1-7 used to calculate the concentration in room
air is difficult to check because we are not given the volume of the room
used. Using only the volume of the inventory and preparation area (that
3 was sealad from Figures 4 .2-1 and 4.2-3) of 1.80 x lo1' ca , we arrive at
total alpha concentrations of about 1 x 10''~ pci/cn3, about 0.2 of those
in Table 6.1-3. Also, the concentrations in Table 6.1-3 are mislabeled;
3 it should be pCi/cn (not di). firthelaore. the W C fraction assupas
the radionuclides are iruoluble form. If they are soluble, the combined
W C would be about 0.17.
However, either concentration results in doses received by the vork force ,,
that are somevhat greater than the values shown in Table 6.1-10. The
concentrations in Table 6.1-3 lead to an annual dose to a worker in the
3 Waste Handling Building (with 22400 m /y inhalation) of 0.43 to 0.54 rem.
depending on solubility assumptions. Yet in Table 6.1-10, it is stated
that the total dose from air contaminants to the 16 waste handlers and 7
radiation control workers in the Waste Handling Building is 0.22 person-
rem. Even with an occupancy factor of 508 for these workers, there
should be a dose of at least 5 person-rem. And if 1008 of the WAC
contamination limit had been used in Table 6.1-4. the dose would have
been at least 10 person-rem. Is it intended that the doses be average
individual doses? (Note our comment on Table U.2-1.)
Furthermore, it is questionabLe if use of a "resuspension fraction" to - calculate airborne concentrations is the w s t appropriate approach.
3 Other concepts such as the 'resuspension factor' (units of Ci/m per
2 Ci/m ) or 'resuspension rate* (units of Ci/sec) have a larger data base
. and usage. Your reference (Sutter) gives data for these other concepfs.
Also, it is probable that w s t exposure will occur to workers near the
contaminated d m , rather than other parts of the Uute Handling Building
vhere the aerosol concentration is much m r e dilute. If a "resuspension
- 6 factor' of 10 /m is used (cornonly accepted value for PuOZ resuspension
in the work place, from page 83 Jones and Pond in 'Surface
Contamination,' Pergcmoon Press. 1964), one calculates that one person-
year in proximity to containers would lead to a 50-year effective dose
commitment of 6 ram.
In summary, we believe that the assumptions and methodology used in
calculating airborne concentrations are questionable, that the calculated
doses presented are much lover than they should be for the assumptions
used. and that worker exposure to airborne contaminants is probably a
much greater fraction of potential occupational exposure than is
concluded in Chapter 6. This entire calculation should be carefully
reviewed.
8. Section 6.1.3.1.1, Pale 6.1-9. Credit is taken for minimizing
personnel radiation exposures by limiting access in the plant. Reference
and document this claim.
9. Section 6.1.3.2.2, Paue 6.1-13. This section discusses use of
Respiratory Protective Equipment. The section does not provide any
detailed information on the S C M equipment to be used. There have been
several recent improvements in the design of such equipment, and the VIPP
should use these improved design criteria. For example, in none of the
VIPP documents is there an indication that S C M equipment to be used vill
be positive pressure demand type, equipment which vill maintain a , .
J'" I ,,
I' . 8 . positive pressure in the face mask (even during high breathing rates),
: ; . - , ; ' , . . _, and provide fresh air on demand. It is only recently that the value of
~!,, :.? ',
*,.;'; '. ,'
. - ,. such positive pressure equipment has been demonstrated. (Some VIPP
documents have referred to Pressure demand SCM's but this could be a
reference to the Negative-pressure respirators, which only provide air on
demand.
10. Section 6.1.3.3. and Table 6.1-7. This section refers to Table 6.1-7
which indicates design contamination zones. Based on this Table. the CH
waste handling area is a Contamination Zone B which has a maximum
contamination of 50 pCi/100 cmL loose alpha. contamination which would be
equal to the WAC contamination limits. It may be difficult. therefore,
to detect contamination at or in excess of the WAC limits if the room in
which the package is located has contamination levels equal to those
limits. Also, it is noted that the Radiation Safety Manual (July 1988)
calls Contamination Zones I. 11, 111. with Zone I having limits of 20-100
2 2 dpm/100 cm (9-45 pCi/100 cm ) removable alpha contamination. The SAR
terminology should be identical to the RSPI and we recommend that the Zone
I limits apply to the CH waste handling area.
11. Section 6.1.3.4.2.1, Page 6.1-16. What is the "derived average source - strength.' and how respiratory protection considered in the result
(i.e., how was the NRC manual on respiratory protection used)? This ,, .
should be clarified in the FSAR.
The use of spacial shielding in separate enclosed area should be based on
1 ram armual dose or 100 mrem/h contact dose rate and AURA
considerations. For example, .n AIARA evaluation right consider worker
dose resulting from installation of temporary shielding as compared to
worker dose from moving drum to a specified location. Other AURA
consideratioru might be potontial contamination problems, existing
radioactive inventories in shielded areas, or worker activity in
controlled areas. In other words, the criteria is too simplistic and
could result in a greater dose to workers rather than less.
12. Section 6.1.3.4.2.4. Page 6.1-19 to 6.1-21. This section fails to
provide information on the basis for the underground shielding for RH-TRU
canisters. For example, in calculating the shielding provided for
emplaced canisters, vas consideration given to those canisters vhich may
have dose rates up to 1000 rem/hr? Information should be added to the
FSAR to indicate the criteria used for shielding of emplaced canisters.
13. Section 6.1.3.5, Page 6.1-21. 'The design and operation of ventilation l i
I- ,$ ,:-~ I systems assure that doses to personnel and the general public . . . are 3, . ,', 3 .
: 9 :. belov limits specified in the appropriate regulatory guidance." Please
. f, .;..;. i 4 ,.<r ,
..-, include the data used. an illustration of the methods and the assumptions
used for calculations, and results.
- 14. Section 6.1.3.5.1, Page 6.1-22. Please provide more detailed
information or data to support the AURA assumptions of this section.
15. Section 6.1.3.6, Page 6.1-26. This section refers to Figure 6.1-13
vhich consists of 11 sheets, none of vhich is legible. Please provide
legible copies.
16. Section 6.1.3.6, Page 6.1-26. The Continuow Air Uonitor (CAM)
description here no longer correctly describes the CAMS (nov alpha CAMS).
What is needed is a figure and detailed description vhich fully explains
the marriage of the NRC and Ebarline equipment added later. A proper
description of a11 these instruments vould involve a statement - really a comiment - to a certain MC-hr response capability in a specified
radon/thoron daughter background. A step in this direction is provided - later for the Pu CAHS. but it still is flawed. There should be an
estimation of performance of the alpha and beta W s in above-ground
applications. and another in the salt dust-laden belov-ground
applications. It is likely that the beta-gamma W s as engineered will
not be adequate in either case. The Pu CAMS oay be adequate in the above-
ground applications. It is difficult to know how the area Radiation
Monitors should be required to be specified, except in terms of the
ability to detect a gamma field producing a given exposure rate at 1
meter from a remote position typical of the installation.
Section 6.1.3.6.2, Paae 6.1-28. Reference to the "posting
requirements" of 5480.lB is not an adequate way to reference WLn
performance requirements. Continued reference to *alpha equivalent" .-
detectors is probably incorrect. If not, this needs some clarification.
l'he concept of airborne radioactivity monitoring is simplistic and not in
accordance with the intent of DOE Order 5680.11. As an example,
NUREC/CR-6033 and NRC Regulatory Guide 8.21 suggest that air samples
should be collected that are representative of the air in the worker's
breathing zone. Sampling of radioactivity shall be in areas known to be
of a greater concentration than that of the worker's breathing zone.
This section should be revised to reflect this fundamental concept.
18. Section 6.1.3.6.2. P a ~ e 6.1-29. On what basis ia an alarm setting at
10 HPC-hrs an acceptable perfonmance standard for W? It should not be
larger than 4 WG-hrs (as is stated later in the chapter). The
underground CAns might have a severe problem meeting G MPC-hrs in a
reasonable radon/thoron background and outfitted with a long transport
line. The FSAR should specify exactly (based on appropriate tests) what
MPC-hr performance can be expected from both types of CAns, in both types
of settings.
References to the CAns "operating" in "expected" environmental conditions _I --
hardly seems a satisfactory specification, particularly with reference to
background suppression, for both alpha and beta-g- W. Furthermore.
if the internal aerosol transport system of the CAM heads has not been
changed to match TAMT design specifications, then from an aerosol
sampling perspective the CAHs in the UHB and below ground are
substandard, and would not be expected to meet reasonable response
standards under accident conditions.
*Calibration" of the W s with plated sources. as stated here. can really
only provide an indication that detector response continues to be the
same as before. With an HCA-based count collection and analysis system
and microprocessor based algorithm for background suppression, a nore
exacting calibration should be done, and explained here. The periodic
demonstration of the proper functioning of such a sophisticated system
will require careful development.
19. Section 6.1.3.6.3. The SAR states that no credible criticality hazard
exists. References please.
20. Section 6.1.4, Page 6.1-30. Uore current data for the waste forms for
VIPP m y be found in revision 3 of reference 22.
21. Section 6.1.4.1. Pane 6.1-30. The design objective should be less than
1 rem/year/person and A m .
22. Section 6.1.4.2. P a ~ e 6.1-31 and Table 6.1-8. The external doses
calculated for routine operations use the February 1985 report. UTSD-TIE-
009, as a basis for calculating external doses received by workers. lhis
report uses older inventory data and time-motion study based on TRUPACT-I
and drums being packaged as six-packs. The UP0 has a July 1988 report,
DOEflIPP 88-012. which uses actual time and wtion data obtained in a
June 1988 Preoperational Checkout that was specific to TRUPACT-I1 and
used the latest inventory data. The later report estimates annual / F!~, , 1 % : . ': external radiation doses from CH-TRU of 13.7 person-rem for normal y,' !.%. ; : \. b<\?, %:, +!
.., *.! .. , '* operation and 0.64 person-sen for "off-normal" waste handling, compared -.. . '
to 9.19 and 0.132 person-rem in Chapter 6. Although EEC has not
critically reviewed WE/WIPP 88-012. the methodology is more specific.
contains real &ta, and uses newer inventories. We believe it should be
used in place of WSD-THE-009.
23. Section 6.1.4.3.2. Paper 6.1-33, 6.1-34. The whole analysis of the
handling of damaged drums seems to presuppose that the workers instantly
recognize that drums or boxes are damaged before they could possibly be
exposed, and don respirators. The experiences recently at the SUEPP
facility at INEL involving a damaged drum show ochervise. Two people - were exposed apparently before anyone donned respirators. Thus, it is
likely that there will be some delay in discovery of damage unless
TRIPACT itself was visibly damaged,
There are several additional questions or observations on this scenario:
a, The origin of the number 66 for contaminated seven-packs is not clear.
From Table 6.1-15 one would expect 36.000/7 - 6860 seven-packs/year and 1% of this would be 49 seven-packs.
. , c : b. From the information given here, the estimate of 1.0 mren/hr seems . >. 6' .I. .. , ,
5, ? : 7 i low. A 6 mrem/hr dose rate at 4 inches would be about 1.2 mrem/hr at Pk.,., ... .'
2 feet, and an exposure to two drums would be about 2.4 mrem/hr
during hands-on operations. Furthermore, WE/UZPP 88-012 uses
average surface dose rates of 16 mrem/hr for drums and 5 mrem/hr for
boxes.
c. The fraction of the waste assumed to be aerosolized and respirable
(FMR) is 10.~. If it is intended that this fraction apply to all
drups in the seven-pack, then the FA6R is equivalent to 7 x loq5 of
one drum. In contrast, the accident scenrrios in Chapter 7 have F M R
values for one drum of 1.25 x (C2) and 2.23 x lo-' (C3) . Is
this consistent? Should an average overpacking and decontamination
operation release as high or higher FA6R as the accidents?
d. We believe the assumption of release of one average drum's
contents is sufficiently conservative. But this is not a negligible
release and it is not clear hov this could lead to a dose as low as
0.04 person-ram. Consider the folloving:
-45-
A
(1) amount released from an average drum (11.9 PE-CI) would be 1.19
(2) this vould be released in 6 volumes of Overpack and
Decontamination ?.oom air, or 1.98 x PE-Ci/volume;
(3) the volume of air (from Figures 4.2-2 and 4.2-3) seems to be
3 9 about 130.000 ft (3.6 x 10 cc);
(4) thus the apparent concentration in air is 5.5 x 10-15 PE-Ci/cc;
(5) a worker would inhale about 6.6 x lo-" PE-Ci/hr (after alloving
for a decontamination factor of from use of respirators);'
(6) for 132 person-hours per year this vould be an intake of 8.73 x
PE--ci/year;
, . , ..,, . .?"',* . (7) for a 50 year committed effective dose equivalent (CEDE) of 430 ; ;, ::.
, , .$!, +q>., k5 ; rem/pCi. this becomes 3.8 person-rem. \, %?e, ".'. ~.,.: -
3 . ?*I ..i .&,%' . . This calculation needs to be carefully rechecked since the results
are potentially significant.
e. We checked the doses in Table 6.1-9 for RH-TRU casks decontamination
vithin 58 and believe the assumptions are conservative.
24. Section 6.1.5.2. Pare 6.1-37. This section refers to Figure 6.1-13.
which should be Figure 6.1-14.
25. Section 6.1.5.4.5, Page 6.1-49. This section refers to "pressure
demand self-contaminated breathing apparatus." Please note Comment 4
(ibow. WIPP should use only POSITIVE pressure demand SCM's.
- 26. Section 6.1.5.6.6. Paae 6.1-50. The last paragraph refers to VIPP 12-
006, "Respiratory Protection Equipment." Reference 31. EEG has been
unable to locate such a report. It probably should refer to VIPP 12-106.
"Respiratory Protection Program."
27. Section 6.1.5.5.1, Page 6.1-51. The first paragraph on this page
states that "Radiation contamination survey stations are indicated on
Figures 6.1-1 through 6.1-6." There are no survey stations indicated on
these figures. except for the underground area (Figure 6.1-5).
28. Section 6.1.5.5.8, Page 6.1-53. What is the criteria for considering
clothing contaminated? The level of contamination in anti-contamination
clothing is important in minimizing skin dose and as an indicator of
contamination problems. The VPO Radiation Safety Hanual does not provide
for hov contaminated anti-contamination clothing will be handled. It /-" "
P"Q, should be made clear whether anti-contaminated clothing will be r 1 8 ~ . I
!:4 b & ,. disposable clothing or laundered. Information on the processing of anti- ?, $,
I h
contaminated clothing should be added to the Radiation Safety Manual.
28. Section 6.1.6.1.1. Page 6.1-56 and Table 6.1-13. When using 1001 of
the WAC surface contamination limit and the other assumptions in this
chapter and x/Q value of 1 x 10" s/m3 at the rite boundary, there
- 7 results a maximum adult dose of 6.2 x 10 rem/y from resuspension of
surface contamination. Although this is a significant increase from the
value presented in Table 6.1-13, it is still a nagligible dose.
29. Section 6.1.6.3, Paae 6.1-63 t h r o u ~ h 6.1-65. In our comments on - Amendment 9 of the SAR. ve noted tha t Table 6.6-5 (now Table 6.1-13) did
not provide data on doses from the several ingest ion pathways, nor does
it compare these doses with those from pathvays vhich d i r e c t l y involve
the plume as claimed on (Page 6.1-66). The UP0 response s t a t e d that t h i s
sect ion is being reevaluated fo r the FSAR to be cons is ten t with the
previous discussion. It appears tha t with only one minor exception. t h i s
sect ion is ident ical t o the previous one, therefore there appears to have
been no response to our e a r l i e r connent.
30. Section 6.1.6.4.2. Pages 6.1-68 ChroupJI 6.1-70. This sec t ion contained
no changes from Amendment 9 of the SAR. Therefore it vas t o t a l l y
unresponsive t o our previous connents on Amendment 9 a s contained in our
l e t t e r t o the UP0 of December 16. 1986. I t a l s o d id not r e f l e c t a peer ,
reviev of the plans fo r s tack monitoring a t WIPP he ld on November 14,
i a. \
1986. i n Santa Fe. The e n t i r e discussion of the e f f l u e n t monitoring \, '
probe design and operation here is writ ten v i t h l i t t l e o r no resemblance
t o the r e a l i t y of the present stack monitoring system and f a c i l i t y
monitoring system, vhich vas t o be completely redesigned fol loving
extensive EEC reviev. Xore specif ical ly:
a . A t present it appears t h a t the alpha CAns i n the s t ack systems have a
vastly improved aerosol t ransport systan of T M design vhen compared
to the f a c i l i t y monitoring systems b u i l t around an Eberline supplied
alpha detector apparently replacin8 the L X-ray de tec tor i n an
othervise unmodified NRC CAM head. Neither hare nor i n the - discussion of the other probes is there any mention of aerosol
transport capabi l i t i es .
4 8 -
b. The shrouded Texas A & I4 University (TAWU) stack monitoring probe is
not designed to be an isokinetic probe as stated here, for example,
the flow rate is kept constant, so there is no "sample flow rate
controller" associated with the TAEN probes, as stated here. A more
nearly correct description is found later in Chapter 6.
c. There is no discussion at all of the expected response time of the
effluent monitoring system, whether triggered by a CAM underground.
or in the stack. This is a significant omission of an assessment of
accidental releases to the environment.
31. Section 6.1.6.4.2. P a ~ e 6.1-69. The statement that the sampling period
and volume for the FAS are "maximized to provide a reasonable lover limit
of detection" is unacceptably vague. There are limits on time and sample
rate for each FAS setting set by the conditions of sampling and the
choice of filter size, filter medium, etc. It should be stated the
lover limit of detection will be for the major TRU radionuclides from the
various FAS locations.
,.-.,... ~, , /,.' ., 3 p%- ! , ,a ,. 32. -Section 6.1.6.4.2, Page 6.1-70. With regard to the response
: I ;
1, . . , , . characterization of the beta CAN (11 cpm after G hrs at 60 lpm), it is 1 ? : . ,
. r; .. almost certainly true that this level of activity could be seen
after a relatively short sampling period (about 1 hr) in the beta-gamma
noise of the radon-thoron background on the filter. As vas pointed out
in EEG-38, the background sensing GI4 tube is not sensing the appropriate
background for this instrument. Bared on the stated sampling rate and
~ P c , using the filter activity equation, the implied efficiency of the A
system at the stated 11 cpm is 341, which is reasonable for a low-
background GI4 detector counting a Sr-90 standard source. But that is
totally inappropriate as a measure of the LID for the system with an
equilibrium load of radon and thoron daughters on the filter. The FSAR
should specify the LID of both the beta-gamma and alpha CAns under
realistic background conditions.
With regard to the alpha CAM response, it is quite adequate to specify an
WC-hr response of 4 WC-hrs for Pu-239. However, it is totally
unacceptable to add the proviso: '. . . when no ocher radionuclides are present." There is no basis for such a provision in ME Orders. ANSI
Standards. or any other authority that is known to EEG. There will
I - ,' always be background radionuclides present on the filter during I . \ ,
monitoring. The FSAR should state the MPC-hr response without
qualification.
33. Section 6.1.6.4.3. Page 6.1-74. The sumary of the preoperational
program in Table 6.1-15 should include data collected by the EEG.
34. Section 6.1.6.4.7. Page 6.1-77. As is evident throughout Chapter 6.
there is no response to EEC'r coments on Amendnent 9 of the SAR. The
EEG had recornended that mention be made in this-section of the agreement
to conduct intercomparison and verification of environmental data with
the State of New Uexico as provided in the Consultation and Cooperation
Agreement. In response, the UPO committed to .Appropriate words will be -
included in the EAR, referring to the arrangement outlined in the
Supplemental Stipulated Agreement through which the State and W E
Environmental Programs are coordinated and through which the State is to
provide independent environmental monitoring verification services to
WE."
35. References for Section 6.1, Pages 6.1-79 through 6.1-84. This
bibliography contains numerous duplications and inaccuracies. For
example. it has five listings of DOE Order 5480.1B. Chapter XI; three
listings of UIPP Radiation Safety Manual; it refers to a 1981 version of
:r
!; W E Order 5480.1B. whereas Chapter * refers to a 1986 version. We h'
. - believe the Chapter 4 listing is correct.
36. Table 6.1-12. Pane 6.1-98. We commented on this Table when it was
presented in Amendment 9 of the SAR. At that time, we were informed by
the UP0 that this data was based on the data in Section 3.1 of the SAR,
but that new source tern data will be used for the FSAR and an example
calculation will be provided for clarification as to how the data in
Table 6.1-12 was derived. The data in Table 6.1-12 is unchanged from the
previous Table. and no example was provided.
37. Section 6.2. P a ~ e 6.2-1. The material on non-radiological surveillance
programs, relevant to agreements, laws and regulations, is a new addition
to the SAR. Its inclusion is desirable. However, the amount of detail
is inadequate. Three omissions are noted in the Section.
38. Table 6.2-1, Page 6.2-8 through 6.2-10. The Resource Conservation and
Recovery Act is appropriately noted in tho Table. However, the extensive
effort and problems concerned with obtaining a RCRA permit for disposal - of the hazardous vaste component contained in much of the TRU waste is
novhere discussed or evaluated. This is an important effort which is
necessary for the FSAR.
39. Table 6.2-2, P a p 6.2-11. List the contract establishing 3EC in 1978.
Same for Table 6.2-3, Page 6.2-16.
60. Table 6.2-3. The Second Uodification to the CQE/State Agreement for
Consultation and Cooperation was signed August 1. 1987, and should be
listed and suaunarized in this Table.
61. Section 6.3, P a ~ e 6.3-1. The last sentence of this Section refers to
Figure 6.3-1. This figure was not included'in EEC's copy of Chapter 6. - 42. Section 6.3.1.1, Page 6.3-2. 6.3-3. This Section should provide
i greater detail for the various training and indoctrination programs
Y> 1, . discussed. or each program described should reference the documents which s, I,# ". provide greater detail.
13. Section 6.3.3. P a m 6.3-5. This Section also should provide more
detailed dercriptiona of the emergency preparedness, or it should
reference other sections of the FSAR or other documents where more
information is available.
U. Section 6.3.3. Page 6.3-6. 2nd Paragraph. The discussion of emergency
medical technicians should indicate tha number of EMT's per LOO - employees.
-52-
6 5 . Section 6 . 3 . 6 . Page 6 . 3 - 7 . Please provide appropriate references to
the Emergency Preparedness Plan.
CHAPTER 7
Accident Analysis
1. There does not appear to be a "Design Basis Accident" identified in this
Chapter. The category "limiting incident" may be considered by the WIPP
Project Office to be an acceptable alternative to the D M (both have an
associated probability such that they are credible, but not expected to
happen over the life time of the project).
The requirements for identification and discussion of D M s are set forth
in W E Order 6430.1, Chapter 1. Both internal and external (e.g.
earthquakej initiating events must be considered. It is not apparent
from the text of this Chapter that a D M assessment has been performed.
It is strongly recommended that such an assessment be performed and 4
summarized in Chapter 7. It should meet the requirements of 6630.1 and
the guidelines of related documents such as "A Guide to Radiological
Accident Consideration for Siting and Design of DOE Non-reactor Nuclear
Facilities.' U-10296-AC. January 1986. The advantage of the D M
; <' ' .'
assessment is that it provides a perspective for evaluating the ~ h . h
. .. k , " : : . . a>, , >,, ., ,, 3 : . t i \ ' significance of the potential mitigators vhich are presently described as
. 't. ,.G -., - ~' limiting impacts to inconsequential outcomes. kloreover, the D M
assessment would reflect conditions vhich might enable structures,
systems and components to be evaluated with respect to their importance
to safety. In a suitable D M assessment. degraded performance of a11
mitigators must be considered, and a release of radioactivity described
associated with the D M vhich causes radiological exposure. This release
event must be shown to result in doses which are not in excess of the
guideline doses in DOE Or&r 6630.1. Chapter I. or associated guidance.
-54-
Another problem in not performing a rigorous DBA analysis is that all
possible failure modes are not systematically considered. The following
two scenarios vere apparently not considered:
a. The contamination of the underground by releases from several
accident scenarios in the CH-TRU portion of the VHB from the
ventilation air flov down the Waste Handling Shaft;
' 8'3: I ( ,,:. . ,A. ic' ;\.
r ! b b \ ~ , , , . . . , . b. Any scenario involving contamination of the radiation control area, . .,
or carrying contamination off-site by workers, visitors, or
equipment. This is perhaps the most likely pathway for radionuclide
contamination from the WIPP Site to reach surrounding communities.
It has been a problem in various nuclear facilities over the years
and considering the difficulty in detecting alpha contamination, is a
potential problem at WIPP. There is no indication in the SAR that
this threat has been adequately addressed by the VPO.
2. There are significant numerical errors and inconsistencies in this
chapter, and serioua questions are raised in the co-nts that follow.
concerning the appropriations of the assumptions and methodologies used.
3. Section 7.2.1. Page 7.2-1. All of the significant CH-TXU accidents are
considered moderate frequency. which is defined as accidents assumed to
occur once a year or up to 25 times during the project lifetime. Yet all
these accidents are assumed to occur only with an average dtun
(apparently 11.9 PE-Ci). We believe that occupatio~l dose commitments
for all of these moderate frequency accidents should be calculated for
the drum loading with the maximum PE-Ci level. These doses would be
significant. For example, with the proposed 1000 PE-Ci limit (which EEC
is objecting to in our letter of June 2 2 , 1988) the 50 year EDE to a
worker in the C2 accident would be about 420 rem and 15 rems would be
delivered in the first year for V category vaster. If the waste were Y
category the 50 year EDE and the 1 year EDE would be about 275 rem and 2 4
rem. In either case. these are very undesirable doses that are possible
and should not be ignored in the Chapter 7 tabulations. Furthermore. the
realization that very high occupational doses are possible should affect
the choice of PE-Ci limit as well as the operational health physics
procedures employed.
4 . Page 7 - 3 . Plea-. indicate the basis for the statement that the relative - radionuclide abundance per cubic foot is greater for drums than for any
. of the alternative containers. If this is true, is the probability of r ' i
release necessarily higher for drums in the various accident contexts ! .5.,;2,, : 5 ". , .. \ , ~ . ,. . given the weight and presence, for example, of heavy, hard wastes in I ,,- . .
I\. I. ., .,. . . . . ~ . boxes? It is worth pausing to consider that the most spectacular release
of TRU haa occurred with an accidsnt involving a box m d not a drum.
5. Section 7.2.2.1, Page 7 . 2 - 2 . The second paragraph on this page states
that 'three' individuals were considered in the analysis. In Section
7.2.2.1.2, second paragraph. .fourg individuals were located within a
single sector. Also Table 7 . 3 - 2 provides dose estimations to a worker,
but fails to show the location of the worker. Was the worker also
considsred to be .outaide the facilities?" A
6. Table 7.2-1, Page 7.2-9. This table from Chapter 3 also has an error in
the Ci/drum of Pu-238. The value should be 1.OE + 01 rather than 1.OE +
02. The PE-Ci should be 11.9 rather than 8. This will affect most other
calculations.
7. Section 7.3.1. Page 7.3-1. This is a new scenario necessitated by the
changed design for TRUPACT-I1 which requires unloading outside of the
_I- ' .. .. ,.: .1 u .
Waste Handling Building. Any release would be to the site and the / :, , 1 - 6.. ..
? ;;$ $,, : ,=$ ., environment from a ground level release without HEPA filtration. It is t, g! s;,>.
.I L b important that this scenario be completed and releases and inhalation '-%." .>. . .
values calculated.
8. Section 7.3.1, Page 7.3-2 throu~h 7.34. This accident assumes that a
bundle (7 drums) are dropped from a fork lift. Also it assumes that the
drop "causes the lids of the drums to be knocked off and an inner plastic
liner to tear.' It is also assumed that 251 of 'the drum contents (8 PE-
Ci) is spilled, and 51 of the total activity is contained in the allowed
one weight-percent.' Therefore these assumptions imply that the total
activity released is
(8 x 7) X .25 X .05 - . 7 PE-Ci
If . W 1 is resuspended this results in 7E - 04 being released, a factor of 7 higher than indicated. Perhaps it was intended to state that the
lid of only one drum fell off, releasing 251 of its contents.
Regardless, this is an excellent candidate for the above-ground facility
D M if analyzed properly. Unfortunately, the opportunity is not
adequately constdared. First, it is a s s w d that a11 workers who could
potentially be exposed in the immediate vicinity of the accident - instantaneously leave, so there is no exposure. In effect lOOI credit is
taken for safety features of the facility and equipment, and for vorker
training. This is not valid unless the circumstances and consequences of
this accident are considered in more detail. The exposure "alloved" is
that to a vorker located in a remote location (and this cannot be
evaluated since Appendix 7B is not complete). Uhy is it incredible
- 6 (probability 10 ) that a worker might not instantly leave the accident
scene? Vhat if he is injured or stunned by the falling drums? What if
he is uncertain about vhether or not a release actually happened? Uhat
if he is trapped momentarily by the equipment and drums?
Although, the scenario description did not state residency time for the
nearby vorker it was estimated that it vould take about 25 seconds for - the *cloud" to travel 20 feet and about 1% minutes of breathing to obtain
the stated inhalation quantity. The forklift operator. at about 5 feet
.- distance, would get the 'cloud' in about 6 seconds and in the next 10
$ ,* 1 I$*
~econds vould inhale about 4.5 x 10.' PE-Ci (7 times that of the worker
' 1' in a 2 minute residency time). Ve believe the assumption is non- " 1
conservative and that a dose to the forklift operator should be
calculated. Incidentally, we note that the 5/86 version of this section
used a .cloudg velocity of 5 ft/min. vhich is only 1/10 of that listed
her..
9. Table 7.3-1. Page 7.3-14. This table has several serioru errors and
inconsistencies. For example. it is noted that the numbers in both 50
year Effective Dose Commitment columns are unchmged from the Kay 1986 -
draft FSAR even though there are substantive changes in the PE-Ci values
released or inhaled for most scenarios. As an illustration. the 6.5 x
PE-Ci inhaled in the C2 accident (using 520 rem/uCi inhaled) the 50
year Effective Uhole Body Dose would be 3.4 ren. If this is further
corrected by using the correct 11.9 PE-Ci average per drum (rather than
8.0) the dose would be 5.0 rem. Also, the 50 year EBD for R4 is not
consistent with that for C3.
10. Table 7.3-2, Page 7.3-15. This table has errors similar to those in
Table 7.3-1. Although the intake for the scenarios is significantly
different in the 1986 draft. the doses in the Table have not been
changed. Accident C3 uses the C2 model to relate resuspended
radioactivity to inhaled radioactivity. With 1.8 times the resuspended
activity the amount inhaled would be 1.2 x PE-Ci. This would result
in an occupational dose of 6.1 reo in Table 7.3-2. With the use of an
11.9 PE-Ci per d w the dose vould be 9.0 ram.
In scenario C4 the calculations check although its not clear why 15
seconds inhalation time vas used h e n the cloud volume vould pass a point
in slightly over 2 reconda. The occupationrl dose comitment in Table > .. 'i 7.3-2 vould be 3.2 rem with an 8.0 PE-Ci d m and 4.8 ram vith an 11.9
PE-Ci d m .
When the C6 scenario is w&led as the C4 scenario the vorker would
inhale 1.1 x 10.~ PE-Ci and the dose vould b. 5.6 ren for an 8.0 PE-Ci
d w and 8.4 re0 for an 11.9 PE-Ci drum.
There are two separate observations on the R4 scenario. One is that if
the maximum canister had 1000 PE-Ci of TRU with the same relative
composition show in Table 3.1-5 there would be 17.3 Ci/l of TRU only.
Thus it would not be possible to have the 17.1 Ci/l of fission and
activation products in the same canister and remain under the 23 Ci/l
maximum concentration permitted by agreement with the State of New
Mexico. See also the preceding c o m n t on Table 7.3-1, which relates to
the R4 scenario.
11. Section 7.3.2. P a m 7.3-7. Accident C7. The first paragraph of this
section refers to WE-AL Order 5484.U. There is no such reference in
Chapter 7. The only W E Order referred to is WE-AL Order 5481.11). on
- Page 7.1-2. Perhaps this is the Order intended here. See also a
r y - reference to DOE-AL Order 5481.U in the first paragraph on Page 7.3-8. - As discussed in Comment 1 above, EEG does not agree that this accident
fulfills the D M assessment requirements of DOE Order 6430.1.
12. Section 7.3.2. Pare 7 .3 -8 . Accident C9. The requirements for diesel
vehicle safety were upgraded years ago folloving discussions vith EEG, so
t h t this scenario could be considered incredible. Please provide the
procedures to bo used to insure that these upgraded requirements are
followed throughout the lifetime of the project.
13. Section 7.3.2. Page 7.3-9. Accident C10. This accident has some
conservative assumptions, however, the last sentence in this description
assumes that no worker is located downstream of tho fire. This does not - seem to be conservative. What is the basis for such an assumption? Note
t h a t Accident CG assumes tha t a vorker i s dovnstream vhen the accident
occurs. A nev employee o r a vorker excited in the presence of smoke
could conceivably v io la te established policy and be located dovnstream
from the f i r e . Also, i s there available experimental evidence t o support
such a large depletion i n a mine environment? Reference G does not
provide such evidence.
This accident becomes the l imiting scenario for o f f - s i t e re leases . I t is
noted tha t the postulated release from a 1000 P E - C i drum is calculated to
de l iver a 50 year e f fec t ive dose commitment of 1 . 4 ram. This intake
vould a lso r e su l t i n a one year dose of 0.51 rem i f the release is i n
soluble form o r 0.67 re= i f it is insoluble. Thus the UIPP system
components pertinent t o t h i s scenario vould be c l a s s i f i ed as 'important
t o safety' i f t h i s vere a comercia1 repository t o vhich 10 CFR 60
applied. Important t o s a fe ty items require special qual i ty assurance
procedures vhich are not presently required of UIPP items.
~ l s o , vh i le the choice of a 1000 PE-Ci drum i n t h i s scenario is very : .. .> . . 2 ,
L !
' . conservative, the x/Q value used is not. I f maximum possible x/Q values
vere used along with appropriate e f fec t ive s tack height correction
fac tors one could ca lcu la te a dose greater than 0.5 rem from a 30 PE-Ci
drum. So, such doses a r e c lear ly possible.
14. Section 7.3.2, Page 7.3-11, Accident Rb. There appears t o be r s l i g h t
e r r o r i n the calculated f i s s ion and act ivat ion products released for t h i s
accident. The r e su l t should be 7.8 x lo-' C i .
CHAPTER 8
Long Tern Waste Isolation Assessment
The Long-Term Waste Isolation Assessment in Chapter 8 . Amendment 9 . contained
159 pages of text, figures, and tables. EEG was in basic agreement vith the - format presented in Amendment 9, although ve responded vith over 10 pages of
questions and comments on details of the assessment. Nov, in the Draft FSAR,
all of these scenarios and evaluations are omitted and replaced vith 9 pages
of general discussion on how performance assessment vill be conducted.
The deletion of the analyses presented in Amendment 9 and failure to provide
consequence analyses is unacceptable to EEG. We understand that DOE is not
prepared to do the Performance Assessment nov and have never considered
Chapter 8 to be the Performance Assessment. Hovever, the SAR is supposed to - analyze the ability of the project to be operated safely in all areas
including long-term vaste isolation. Hov can EEG and others reviewing the SAR
conclude that the facility is adequate for long-term vaste isolation when the
SAR is completely silent on this?
The inclusion of the compliance strategy in Chapter 8 along vith the long-term
vaste isolation assessment is desirable. Hwever. the strategy presented is
very brief and largely avoids mentioning specific &tes or other details.
This compliance schedule should be included in the FSAR. Also, on page 8.1-2,
the last sentence refers to *&ta collection and development during the Pilot
Plant Phase will be required to complete the predictive models being used for
the long-term performance assessment." Please describe in more detail the
data and predictive models referred to. Other comments on specific sections
of this Chapter are provided below.
1. Section 8.1.1, Page 8.1-3. As stated on Page 1.b of ME-UIPP 86-013. the
Compliance Strategy vill include Sensitivity AND UNCERTAINTY (emphasis
added) analysis. The third item on this page failed to include the
uncertainty aspect of the analysis. It is particularly important that
the DOE analyze the parameter uncertainties, i.e.. uncertainties about
. . the numerical values in or resulting from data. and uncertainties in the
conceptual model and its mathematical representation.
2. Section 8.1.1. Page 8.1-3. This section indicates that the final set of
scenarios to be modeled for inclusion in the CCDF will be chosen from 73
scenarios examined during the PA. Because of lack of data on
probabilities from actual experience, ve urge that the W E have a Peer
Review of the scenarios and utilize the recommendations resulting from
such review for the selection of the final set. EEG should participate
in such Peer Review. Although the last paragraph on Page 8.1-7 indicates
that a Peer Review Panel will evaluate the PA. there is no indication of
the timing of such review or of what aspects will be reviewed by the
Panel.
3. Section 8.1.1, Page 8.1-3. The last bullet on this page concerning
human intrusion modeling should include Castile brine resenroirs. (See
EEG reports EE-11 and EEG-15.)
T--
6. Section 8.1.3. Pages 8.1-6 to 8.1-6. This section completely ignores
human intrusion scenarios that result in waste being brou&t to the
surface (which is also the accessible environment). This is likely to be
the most serious single pathway.
5. Section 8.1.6, Page 8.1-7. The last paragraph of this section states
that 'a Peer Review Panel will provide assurance to the WE/WPO that the
performance assessment is well conceived and being carried out with
professional competence, and so that scientists and state officials can
be assured that DOE'S ultimate conclusions as to the suitability of WIPP
as a repository are credible." In the past, the DOE/WPO has refused to
include the State Environmental Evaluation Croup (EEG) as a participant
or observer of such Peer Review Panels. Until EEG is routinely included. -.
and given reasonable advance notice of such reviews, with agenda and
background information on proposals, the ultimate conclusions are not
likely to be considered credible by state officials, nor by other
scientists having no organizational ties to DOE. If the DOE/UPO plans to
establish a new policy vith respect to these Peer Wviev Panels, such
that they will iaelude full participation by EEC, it should be indicated
in this section.
CHAPTER 9
Conduct of Operations
1 . Section 9.2. page 9.2-1. The f i r s t paragraph indicates tha t "formal
t e s t procedures are prepared for complex and c r i t i c a l systems as
discussed in the Operations Program Plan." The FSAR should l i s t those
complex and c r i t i c a l systems, and should provide more d e t a i l on the t e s t
" ... ,.. 2. Sections 9.2.1 throuph 9.2.2, Pages 9.2-1. 9.2-2. These sect ions should
provide more de t a i l on the acceptance t e s t i ng programs. To what extent
a r e equipment and systems tes ted at the factory? What f racr ion a re found
t o be defective? What f ract ion of systems checked a t the UIPP S i t e a r e
found t o be defective? Where a r e such t e s t i ng records maintained? What
documents contain the tes t ing procedures? What documents contain the
procedures for handling nonconformance? This type of addit ional d e t a i l
should be providtd or t h i s sect ion should reference vhere such d e t a i l is
documented.
3. Section 9.3, Pam 9.3-3. The first paragraph on page 9.3-3 re fe rs
incorrectly t o 'regulations n e e r e d 48. 49. and 7 4 i n 30 CFR 57 . . . . "
The correct reference should be t o "par ts 48. 49, 57 . and 74 i n T i t l e 30
of the CFR.' hovever. Par t 74 is applicable t o coa l mines only.
4. Section 9.3. Pam 9.3-6. Reference 2 is a l s o incorrect i n t ha t it
re fe rs to only one Pa r t ( i . e . . 5 7 ) of T i t l e 30. vhereas it should r e f e r
t o T i t l e 30. CFR, vi thout specifying par t s .
- 5. Section 9.6.1. Paae 9.4-1. Please provide a reference to the formal
written operating p~rocedures.
6 Section 9.4.2. Pa~e9.6-2. The third paragraph of this page refers to
30 CFR 57, whereas a more appropriate reference is 30 CFR 48.
7. Section 9.4.4. Page 9.4-2, 9.4-3. The last paragraph of 9.4-2 refers to
"abnormal' occurrences. To be consistent with the terminology of other
DOE and UIPP documt!nts, the word "unusual" should be used instead of 4'
t...,:. "abnormal.' For example, see the definition of "unusual occurrence" in f " .;
V.'" ., ,.* ,. , . i ,,
DOE order 5484.2. The word "unusual* should also be uaed,in the first i ". 'G, ,;.:.
-. _ . and last paragraphs of page 9.4-3 instead of "abnormal" and "internal:
8. Section 9.4.4.2. Page 9.4-4. The discussion in the first paragraph of
this page should ilnclude a requirement t b t the UIPP operating contractor
shall notify approlpriate state authorities to assist in determining the
impact of the vio1,ation on public health and safety. For example, the
New Nexico/DOE C & C Agreement requires that 'WE shall keep the State
currently and fully advised . . . so that the State m y m&e independent
revieus on public health m d safety concerns relative to UIPP." To
maintain credibility, it should be considered essentid by UW to involve
the State iuthorities at an early stage in the public health evaluation.
Furthermore, the UP0 must meet the public exposure requirements of
40 CER 191. Subpart A. These regulations should be referenced in Section
9.4.4. -
CHAPTER 10
Operations Safety Requirements
1. This chapter lacks sufficient detail to permit EEC to adequately evaluate
the operational safety at WIPP. It begins by defining safety limits but
then states that there are no variables which require safety limits (see
Sections 10.2). However, there are provided sections called
"Surveillance Requirements" for the safety limits. The references in ,-, * " '
., ' Fj i:
some cases are so general as to be of little value. In all instances. [ .:, i . , the references do not provide the complete title, nor do they cite the 5 ;; ,' . .
Y applicable section of the referenced documents. It is recommended that
this Chapter be extensively revised and specific sections be revised as
indicated in the following comments.
2. Section 10.1.1. Pane 10.1-3 through 10.1-4. Please add and define the
following ACRONYMS vhich are used in the Chapter, or should be cited in
the Chapter:
noc - (see page 10.3-9) W A C - WAC - Waste Acceptance Criteria
3. Section 10.2. Page 10.2-1, S a f e ~ Limits. This paragraph states that
there are no measurable process variables vhich could result in an
unfiltered release of radioactive material that exceed the DOE guidelines
at the plant boundary. It vould seem that the MIE Waste Acceptance
Criteria should be construed u such a set of variables. Although
unlikely, it should be considered credible that a waste shipment could be
received which has external contamination well in excess of the WAC. A
Furthermore, the DOE radiation exposure limits should be construed as
safety limits or me,asurable process variables.
4, Section 10.3. Page 10.3-2. It is recommended that the vord "activities"
in Item 3 be amended to 'activities, equipment or services".
5. Section 10.3.1.3, Page 10.3-3. In the first and second paragraph, the
language should be amended to preclude, without exception. any operation
vith radioactive material vhen the WLELs and ARMS are nor operating.
These devices must be considered as essential to a central monitoring
system. When not operational, the operations vith radioactive materials
should be suspendecl until these devices, or substitute equipment. are
again operational. Therefore, in the first paragraph, please delete the
words "except during calibration and change out.' In the second
paragraph delete .nnless otharviae noted' and '(within a o m hour
l i m i t . The 'Exception" on page 10.3-4 is quite reasonable and proper.
_<.-._ if . a, hovever it is recommended that the vords 'continuously wnitored by HP
ye ;:!,
personnel . . . . be changed to 'continuously monitored with equivalent
equipwnt..
6. Section 10.3.1.3. Page 10.3-3. Item F. *SO ft' should bo revised to
"50 ft or less.'
7. Section 10.3.1.4. Page 10.3-4. The language under "Basism is so general
as to be useless. It should be expanded, or replacod, to state the -_ applicable sactioris of specific rtandarda or guides.
8. Section 10.3.1.5, Page 10.3-4. To avoid confusion. please provide
specific and complete references.
9. Section la.3.2.5. Page 10.3-5. The reference "Westinghouse 0 and U
Procedures" is inadequate. Please provide specific references.
10. Section 10.3.3.3, Page 10.3-6. Please revise the mid-page notation to +
pi. state that the HVAC systems for the above ground facilities in the VHB
are independent. The underground ventilation systems do not provide
independent operation in the CH and RH areas.
11. Section 10.3.3.3. Page 10.3-6, Paragraph 2. Item 2 needs to clarify
whether a11 (above ground and underground) ventilation will be shutdovn.
or whether it will only be shutdovn in the affected work area.
12. Section 10.3.3.3, Page 10.3-6. Paranraph 3. Item 3 needs to be amended
to state what minimum air flow is needed when normal operations are
stopped. What air flow is required when HP survcys are being made to
determine the area affected by an unplanned radiological incident or
release, or when an air monitoring instrument is not operable.
13. Section 10.3.3.5. Page 10.3-7. The references should cite the
applicable sectiona of the referenced documents. Also the referenced
documents should be complete, including reference number, etc.
14. Section 10.2.3.4. Pane 10.3-8. The referenced citations should be A
complete and should cite specific sections applicable.
15. Section 10.3.5.1. Pane 10.3-8. The applicability should include
portable fire supprlession devices such as extinguishers, fire hose and
stand pipe equipment.
16. Section 10.3.5.3, Page 10.3-8. This paragraph should be more specific
to include minimal water pressure of sprinkler and hose systems.
17. Sections 10.3.5.4 and 10.3.5.5. Page 10.3-9. Cite specific sections of
the referenced documents and include document numbers.
18. Section 10.4.1. P a m 10.4-1. Please cite specific sections of the UIPP A
operating procedures and mainterunce instructions.
19. Section 10.4.2, Page 10.4-1. Cite specific sections of reference
documents vhich apply.
20. Section 10.4.2. Pare 10.4-1. Paragraphs 1 and 2.. b .nd c. These
paragraphs should be revised to usure that records of each inspection
vill become a part of the central records system.
21. Section 10.4.2. Page 10.4-2.
Paragraph 1 - The local monitoring should be equivalent to the CPlS capability.
Paragraph 2 - Should be revised to assure that the air quality is checked frequently and to document the air quality standards.
Paragraph 4 - Should be titled "Exhaust Filter Building Equipment." and paragraphs ia and b become subparagraphs. Paragraph ib 4 ' '',. , should require a.monthly demonstration of the ability of
5 i:3: < ,.. I? : $ the equipment to shift autoaatically to the filtration ; .,
1 ,' $ .' . . mode.
22. Sections 10.6.6 and 10.i.5. Page 10.4-3. Please include specific
sections of the VIPP Safety Manual which apply to inspection and testing.
Also include the requirement that the records of the testing be included
in the Central Records System. Furthermore, the equipment to be tested
or inspected should include portable fire extinguishers and stand pipe
systems.
23. Section 10.5, Page 10.5-1 through 10.5-3. This section seems
superfluous since it does not include limiting conditions of operations.
It should be edited to be consistent with other sections to include
specific LCO's. or it should be eliminated. Chapter 10 is not the
chapter for descriptions of systems.
24. Section 10.6. Paws 10.6-1 through 10.6-16. This section should be
revised to include the role of the Quality Assurance (QA) program. The
QA manager's role is indirectly referenced in 10.6.5.3. Specific QA
responsibilities should be stated, or referenced, relative to such items
as 10.6.1 Training, 10.6.2 Design and Procurement, 10.6.3 Document
Control. 10.6.4 Auditing, and 10.6.7. 10.6.8. and 10.6.9 on radiological
protection. Quality Assurance is essantial to operational safety.
-71-
A.
25. Section 10.6, Pace K u . The paragraph entitled "Administrative
Controls for Operations* states that "Administrative Controls shall
define the mechanisms and interfaces necessaq (emphasis added) to insure
adherence to the Sa:Cety Limits and LCO's, DOE orders and UIPP Standards
Practices." If this statement is referring to the Administrative
Controls of Section 10.6, these controls do not seem to meet this
definition. The Adrrinistrative Controls of Section 10.6 do nothing more
than refer to other documents in such general terns as to be of minimal
value. Also, in some cases, the "controls" are nothing more than a
restatement of previous sections of Chapter 10. As examples of the
foregoing, Section 10.6.2.3 does not state a "control;' it states that a
control shall be established. Similarly, Section 10.6.3.3, Paragraph 2
states that the operation manager shall establish a control. Section - 10.6.7.3 is a restatement of portions of Section 10.3.1.3.
CHAPTER 11
Quality Assurance
EEG has not reviewed the quality assurance plans and procedures of various
aspects of the project and is, therefore, not in a position to evaluate these
adequately. Quality assurance relative to the waste acceptance criteria and
waste certification at the generator sites have been performed and the
comments are contained in other chapters of draft FSAR.
CHAPTER 12
Decontamination and Decomissioning
While it would be unreasonable to expect DOE to provide detailed analyses of
decomissioning proceduxes at this time. it is reasonable to expect some
indication that the Department is addressing the problems involved as shovn
in the folloving:
1. Section 12.1, Page 1.2.1-1. There is no commitment here to meet the
D & D requirements 'of EPA's Standards for the disposal of TRU vaste,
40 CFR 191, which slhould be included. This section should be expanded to
include specifics en the regulations of 40 CFR 191. as they relate to the
preparation for Dec~ontamination and Decommissioning. For example,
pursuant to DOE'S " A Plan for the Implementation of Assurance - .*
,.."
* ~'i Requirements in Compliance 4 t h the 40 CFR 191.14 at the Waste Isolation ; . ,h\
,.:, ~- : . . : Pilot Plant." WE/WIPP 87-016. the WIPP vill become a disposal facility
. , ,< ,~ . . if the decision is made (5 years after receipt of the first vaste) not to
retrieve the vastes. The UW mrst then begin to implement the VIPP
active institutional controls program as required by 40 CFR 191. Subpart
8, and WE/VIPP 87-106. This program includes four important steps,
vhich are listed on, page 8 of DOE/UIPP 87-016. and should be listed in
the FSAR, Section 12.2. These steps relate to the advance preparation
for Decontaminatiorn and Decomnissioning. Furthernore, this section
states that Chapter 12 is written with the assumption that WXPP is shovn
to be acceptable as; a repository and, therefore, decomissioning
activities begin near the end of the operational life (25 years) of WIPP.
Such an assumption reflects a bias in favor of the acceptability of the
five year pilot phase. If the five year phase is to provide a valid
test, then such a bias should not be condoned. Also, if the
demonstration does indicate that permanent disposal is unacceptable,
there should be available, well-in-advance, a detailed plan for removal
of the waste, decommissioning of the facilities, and decontaminations and
environmental evaluation. This chapter does not provide this degree of
detail. This degree of detail is needed to allow for sufficient
budgeting, agencies and facilities, and to provide a base for post
decoaaissioning activities such as the disposition of the land.
2. Section 12.2. Page 12.1-1. The text states that a11 the waste will be
retrieved-if th& site proves to be unacceptable during the five-year
demonstration period. Then the text states that this Chapter is vritten
with the assumorion that WIPP will be acceptable. Hence, it will not be
necessary to retrieve the waste. Since WE has consistently insisted
that it is necessary to enplace 125.000 d- during this five-year
period. specific plans and consequence analyses should be provided as to
whether the 125,000 drums will be returned to the generating sites, sent -. v to the high level waste repository, left on the surface at WIPP or be
' x
held underground before a new site is selected.
b'
3. Section 12.2.1. 12.4. and 12.5. The term "Decomissioning Plan" in
Chapter 12 includes varying adjectives of "firul" and "detailed." Are
they different?
4. Section 12.3, Page 12.1-2. Since the Department has requested the
Congress to assign exclusive responsibility to W E to prevent future
mining during the post-deco~issioning phase. Section 12.3 should state - vhether the Department plans to reassign those authorities to the BLW.
Department of Interior. State of New Mexico or establish a W E
organization to prevent mining rather than merely rely on records and
documents.
5. Section 12.b. Page I U . Uhile the three goals of closure are
laudable, no indication is presented of how they are to be coqleted.
The text should provi& some indication of plans for active and passive
controls, how long W E will maintain a fence, employ watchmen and
actively prevent mining permits in the area. Also, the various
regulations- and orders that oust be met should be mentioned and
discussed. These include WE Order 5820.U. 40 CFR 91. 40 CFR 265
Subpart G, and NMiUMR 206.C.2.
6. Section 12.5, Page 12.1-5. Post-closure environmental surveillance
within the repository should be determined before the disposal mode
begins in order to allov installation of sensors at the time of
emplacement .
CHAPTER 13
Glossary
1. "Aquifer" - This definition contains a misspelled word "aquiclide", which should be "aquiclude."
2. This glossary should be expanded to include additional terms used in
several chapters and other WIPP documents, such as:
a. Radiation area
b. High radiation area
c. Unusual occurrence